ML20202C058: Difference between revisions

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{{Adams
#REDIRECT [[IR 05000282/1997023]]
| number = ML20202C058
| issue date = 01/30/1998
| title = Insp Repts 50-282/97-23 & 50-306/97-23 on 971203-980113. Violations Noted.Major Areas Inspected:Licensee Operations, Maint,Engineering & Plant Support
| author name =
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
| addressee name =
| addressee affiliation =
| docket = 05000282, 05000306
| license number =
| contact person =
| document report number = 50-282-97-23, 50-306-97-23, NUDOCS 9802120155
| package number = ML20202C041
| document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 27
}}
See also: [[see also::IR 05000282/1997023]]
 
=Text=
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                                    U.S. NUCLEAR REGULATORY COMMISSION
                                                    REGIONlli
                          Docket Nos.          50-282; 50-306
                          License Nos.        DPR-42; DPR-60
                          Report No.            5(. 482/97023(DRP); 50-306/97023(DRP)
                          Licensee:            Northem States Power Company                                .
                          Facility:            Prairie Island Nuclear Generating Plant
                          Location:            1717 Wakonade Drive East
                                              Welch, MN 55089
                          Dates:              December 3,1997, through January 13,1998
                          Inspectors:          S. Ray, Senior Resident inspector
                                              P. Krohn, Resident inspector
                                              S. Thomas, Resident inspector
                          Approved by:        J. W. McCormick-Barger, Chief
                                              Reactor Projects Branch 7
                9002120155 900130
                PDR  ADOCK 05000282
                G                  PDR
                                                                                        - - _ _ _ _ _ - _ _
 
                                      -_____                          __ _ _ - _ _ _ _ - _ _ - _ _ _ _ _ __ -________ - -______-___ _ ___ - _ _.
l-        n e
  o _, s
L
l
                                                        EXECUTIVE SUMMARY
                                          Prairie Island Nuclear Generating Plant, Units 1 & 2
                              NRC Inspection Report No. 50-282/97023(DRP); 50-306/97023(DRP)
              This inspection included aspects of licensee operations, maintenance, engineering, and plant
              support. The report covers a six-week period of resident inspection.
              fioerations
              *      Management expectations and procedures for conduct in the control room, such as those
                      delineating the frequency and completeness of main control board walkdowns, were not
                      always clear, in addition, first line supervisors did not always enforce those procedures
                      that were clear, such as those relating to communications and control room access
                      (Section 01.1),
              o      Unit i startup operations from the refueling outage were generally conducted well with no
                      significant problems. Procedures were followed and operators remained attentive to
                      plant indications during plant mode changes (Section 01.2).
              *      One instance occurred during the Unit 1 startup where an operator did not verify that an
                      annunciator (the ROD AT BOTTOM annunciator) had cleared in a timely manner.
                      Although this error was not safety significant, it emphasized the need for improvements in
                      procedure organization and for further evaluation of procedure use expectations
                      (Section 01.2).
              *      Following the retum to full power operations after the Unit i refueling outage, power was
                      reduced to 5 percent to allow balancing of the main turbine. Control room activities for
                      the power reduction, turbine balancing, and retum to full power were conducted well
                      (Section 01.3).
                                                                                                                                                                                                .
              o      During a walkdown of the Unit 2 containment spray and caustic addition systems, the
                      inspectors found the systems properly lined-up and ready for safeguards operation. No
                      significant material discrepancies or system deficiencies were identified that would
                      prevent either system from performing its intended function (Section O2.1).
              Ma'ntenance
                e      Operators involved in maintenar.ce and surveillance activities displayed a good
                        questioning attitude and appreciation of radiation dose control (Section M1.1).
                *      A good questioning attitude by an operator resulted in identification of an inadequacy in a
                        procedure for main turbine torsional testing. However, the initial review of the operator's
                        concem by engineering was poor, and the concem was not validated until the test was
                        started and equipment did not respond as expected (Section M1.1).
                e      The inspectors identified that a physics testing procedure had not been followed in that
                        the amount of reactor coolant system temperature change called for la the procedure was
                        not accomplished (Section M1.2).
                                                                    2
                                                                                                                                                _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
 
    . w e
  ns
          e      The inspectors identified several minor deficiencies in the surveillance Mures for
                  operational pressure test inspections of the cooling water system (Section M3.1).
            Ennineerino
          e      System engineers were heavily involved with all asper:ts of operations, maintenance, and
l                testing of their systems. The engineers rapidly investigeled any operational
                  abnormatities, took an active role in maintenance and troubleshooting activities, and
                  closely followed alt surveillance testing on the',r systems; however, in one instanca, during
                  turt>lne torsional testing, a system enginser did net provide adequate technical support
                  (Section E2.2).
          *      Recer:< findings by tne engineering orgsnization involving the control room ventilation
                  system and a 10 CFR Part 50, Appendix R issue regarding inadequate separation of
                  pressurizer level cables indicated that thorough design reviews were being conducted
                  and reflected a wi'lingness to Idantify and resolve old jesign and compliance issues
                  (Section E3.1).
          Plant Sucoort
          e      Good involvemeM of radiatk,n protection personnel in job p'- aning and execution in order -
                  to rnaintain low ooses was observed, as exemplified by the involvement of radiation
                  protection personnel with operators during performance of a reactor coolant system
                  integrity test (Sections M1.1 and R1),
          e      An old Appendix R compliance issue involving inadequate separation of pressurizer level
                  cables was identified and rapidly corrected as a result of a proactive lic6nsee initiative
                  (Section F2.1).
                                                              3
              ,                                                              _
                                                                                          -
                                                                                                                _ _ _ _
 
                                                                                                -
                                                                                                                              .
                                                                                                                                -- . - . -
        . - m ,
                                                                                                                                                                  l
%, < -
'
                                                                                                                                                                  l
;
                                                                        Report Details
i                Summary of Plant Status
                Unit 1 was restarted upon completion of a refueling outage on December 13,1997, and the
4                generator was placed on the grid for the first time on December 14. After extensive test 8ng of the
                newly installed turbines, the b,'.it reached full power on December 19. Power on Unit 1 was
                reduced to about 5 percent on January g,1998, and the generator was taken off line in order to                                                  !
*
                accomplish turbine balancing. The generator was placed back on line January 10 and the unit
                retumed to full power on January 11. Unit 2 operated at or near full power for the entire
                inspection period.
!'                                                                      l. Operations
                01      Conduct of Operations                                                                                                                  i
                01.1    General Comments
                  a.    inspection Scoce (71707. g2901)
:                        The inspectors conducted frequent reviews of plant operations. The inspectors
                        performed observations in the control room for extended periods and focused on shift
;
                        tumovers, prejob briefs, communications, control room access control, logkeeping,
                        control boarc monitoring, and general control room decorum. Section 13. "P%nt                                                          ,
;                        Operations," of the Updated Safety Analysis Report (USAR) was reviewed as part of the
-
                        inspection.                                                                                                                            ;
                  b.    Observations and Findinas
'
                        The inspectors noted that shift tumovers were usually good, covering the status of both                                                ;
                                                                                                                                                                ~
                        units, on-going maintenance and evolutions, and other specific instructin1s for the safe
                        operati sn of the units. However, on two occasions operators arrived late for the moming                                              '
'
                        control qum shift tumover briefing, missing significant portions of the information -
;                        presented.
                        The inspectors observed numerous projob briefs, including briefs for int < rated safety
                        injection testing, Unit i reactor startup, Unit i reactor physics testing, and Unit i turbine                                          '
;-                      overspeed and torsional testing. Generally, the briefs were concise, but thorough. The
,                        inspectors noted that the use of formal communications, the slow and cot trolled conduct
                        of the evolution, and reactor safety were common issues stressed in each brief. No
L                        specific discrepancies were noted.
                        The inspectors observed operator commu,,ications during numerous evolutions including
                        both routine and noteroutine operations. While communications were deemed adequate,
'
                        they rer.ged from excellent to poor, depending on the evolution in progress and/or crew
                        ob. served. Specific observations included:
                                                                                                                                                                :
l-                        e        - the consistency with which formal ccmmunications were used varied from crew to
                                    crew;
                                                                                        4
i
.                                                                                                                                                              .
6
  ,..,-o                  ,---,,e        -,w-  --,yw ..,-.v- ,,,- ----m,n,w--nne.,n,--  y - , ma-g-y-, y-n,y,-,ne-w w~,,ow,              . - - , . 7, yp p r
 
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                                                                                                          ,
            e      formal communications tended to be used more frequently during planned
                    evolutions, such as reactor startups, but less frequently during abnormal operaHng
                    situations; and
            e      communications in the plant were generally not ab formal as those in the
                    simulator.
            During tl o extended periods of control room obscrvation, the inspectors monitorea how
            control ri cm access was managed. Section Work Instruction (SWI) 0 2, ' Shift
            Organta tion, Operatica, and Tumover," Revision 38, stated that the 'thift supervisor (SS)
            *shall be responsible for maintaining control of personnel entering the control room" and
            the lead plant equipment and reactor operator (LPE&RO) shall be responsible for
            " granting permission to non-operations personnel for entry Ido the control room."
            Implementation of those aspects of SWI O 2 was poor. Specific discrepancies noted in
            the control of controt room af4:ss were:
            e      on numerot.s occasions, instrument and control technicians entered the control
                    area boundaries within the control room and approached control panelt without
                    first obtaining permission and/or informing the LPE&RO of their reason for doing
                    so; and                                                                                4
            e      on several occasions, personnel entered the control area of the control room and
                    approached control panels wecring hard hats even though there was a si-ln,
                    located at the entry to the control area, which stated that hard hats were not
                    allowed beyond that point.
            The inspectors observed control room operators during the performance of routine log
            taking and control board monitoring. The inspectors considered the Operations Log and
            the individual Unit 1 and Unit 2 Reactor Logs to be an accurate accounting of shift events
            and noted that relevaret shift information was consistently logged. The Inspectors could
            find no specific uperator guidance, nor could any be provided by the poneral
            superintendent of operations, on the frequency that it.e control berds should be " walked
            down." Control board monitoring was considered adequate, but me time between
            walkdowns varied widely from crew to crew. Specific discrepancies noted in control
            board monitoring were:
              o      the continuous monitoring of control boards exhibited by operators participating in
                      graded simulator scenarios was not observed in the main control room;
              e      during one of the inspectors' control room tours, the inspectors noted that over a
                      two hour period of time, the only time the Unit 1 panels were walked down was
                      when the hourly logs were taken;
              e      annunciators were frequently silenced and acknowledgod by a single operator
                      without announcing to other control room personnel what the alarm was and the
                      reasoa for the alarm; and
              e      the inspectors identified tht.t incorrect work order numbers were referred to in one
                      reactor log entry. When identified, th SS corrected the log entry.
                                                          5
 
                                                                    _
        . .. ,
  *yi
'
                      The inspec* ors observed the overall control roor.1 decorum during evolutions which
                      ranged from complex to routine. The inspectors gens, ally categorized the overall
                      atmosphere as relaxed, but professional. The appropriate level of conoom and
                      supervisory oversight was demonstrated during complex and Infrequently performed
                      evo;utions. The SSs and shift managers generally handled administrative matters,
                      leaving the other control room operators free to monitor and t,ontrol 6ach Unit's operation.
                      The inspectors noted that some activities detracted from a professional control room
                      atmosphere, including:
                      o      eating food and/or drinking beverages while , < rating components at a control
                              panel;
;                      e      inappropriate screen savers on the computer monitors in the control room;
                      e      extended discussions of toples not related to the operation of the plant; and
                      e      inappropricte material posted on the walls of the SS's area.
                c.    Conclusions
                      Management expectations and procedures for conduct in the control room, such as those
                      delinteting the frequency and completeness of main control board walkdowns, were not
                      always clear. In addition, first line supervisors did not always enforce those procedures
                      that were clear, such as those relating to communications and control room access.
                      inconsistencies in performance between crews indicated the need for additional guidance
                      and tialning in this area. The discrepancies discussed above did not lead to any unsafe
                      conditior,s or violations of NRC requirements. The plant manager informed the inspectors
                      that a revised section work instruction on control room access and other expectations
                      was being developed.
                01.2 Unit i Retum to.100 Percent Power Operation
,
                a.    inspection Scope (71707)
                      The inspectors observed significant portions of operations leading from a refueling
                        shutdown to 100 percent power operation on Unit 1. Major activities observed included:
                        e      integrated safety injection test;
                        e      transition from Mode 5 to Mode 4 (Cold Shutdown to Intermediate Shutdown);
                        e      drawing a pressurizer bubble;
                        e      transition from Mode 4 to Mode 3 (Intermediate Shutdown to Hot Shutdown);
                        e      reactor stanup;
                        e      reactor physics testing;
                                                                  6
    .. .
                            ..    . . . . .                  .
                                                                      .
 
                                                                                _                _ _ _ _ _ _ _ _ _ _ _ - _ -
      . > . .
  ...<
                e      transition from Mode 3 to 2 to 1 (Hot Shutdown to Hot Standby to Power
                        Operation); and
                e      turbine overspeed and torsional testing.
i
                included in the startup observations was a review of the appropriate USAR sections and
                operating procedures regarding the activities. The inspectors verified that Jpplicablu
                surveillance procedures performed as part of the startup met the requirements of the
                Technical Specifications (TSF).
              b. Observations and Findinat
                For most of the evolutions observed, procedures were properly used and followed.
                Operations personnel demonstrated experience and knowledge during the performance
                of their tasks. Noteworthy comments on specific evolutions are discussed below.
                  *      The inspectors attended the prejob brief and obse ved performance of
                          surveillance procedure SP 1063, * Unit 1 Irstegrated Safety injection Test With a
                          Simulated Loss of Offsite Power," Revision 24, from the control room and
                          emergency diesel generator rooms. The prejob brief was thorough and attended
                          by the plant manager who stressed proper command and control, personnel
                          safety, nuclear safety, and equipment protection.
                  *      The inspectors observed good command, control, and coordination of activities
                          during the integrated safety injection test. The complex test required the
                          coordinated effort of many operations, engineering, and maintenance personnel to
                          establish the required test conditions and monitor system performance as the test
                          was performed,
                  e      The inspectors observed the prejob brief conducted prior to the Unit i reactor
                          startup. Reactivity manegement, expected indications, and personnel roles and
                          responsibilities were discussed. An extra reactor operator (RO) and SS, in
                          addition to the normal crew complement, were assigned to perform the startup.
                          Other plant activities and distractions were kept to a minimum. Nuclear
                          engineering personnel were also present and perform 3d independent verifications
                          of reactivity management as the startup progressed.
                          The inspectors observed the withdrawal of the control banks and dilution to
                          criticality. The SS and RO remained attentive to reactor power ievels and startup
                          rate iridications throughout the reactor startup. However, after control
                          bank A rods had been withdrawn to 129 steps, the RO noticed that tr,e ROD AT
                          BOTTOM annunciator had not cleared. The reactor operator drove bank A rods
                          to O steps and a work order was issued to investigate the cause of the
                          annunciator not clearing at 20 steps as expected. it was deteImined that the
                            pulse to-analog bistable for control bank D had failed causing the ROD AT
                            BOTTOM annunciator to remain energized (the position of rods in all four of the
                            control banks input into the logic for the annunciator).
                                                                7
                      .                                                .
                                                                                          _
 
  . . . .
3i
                        Operating Procedure 1C1.2, " Unit i Startup Procedure," Revision 18, Step 5.5.0,
                        instructed the reactor operator to startup the reactor per Appendix C18, " Appendix
                        - Reactor Startup," Revirion 6. Appendix C1B was intended to be an aid to the
                        RO for conducting the startup and was not required to be "in-hand' during the
                        actual startup evolution since the operator's attention should be focussed on the
                        control board. Step 5.2.2.C of Appendix C18, required th;; RO to verify that the
                        ROD AT BOTTOM annunciator (47013-0407) cleared with bank A control rods at
                        appro41mately 20 steps during the reactor startup. The RO did not perform that
                        verification until the control bank A rods had reached 129 steps.
                        Technical Specification 6.5.A.1 required that detailed written procedures for
                        normal startup of the reactor be prepared and followed. On December 12,1997,
                        Operating Procedure C18, * Appendix Reactor Startup," Revision 6, Step 5.2.2.C,
                        was not followed wnen the RO did not verify that the ROD AT BOTTOM
                          annunciator cleared when rod bank A was withdrawn to approximately 20 steps.
                          The RO later identified that the annunciator h::J not cleared when he stopped
                          moving bank A rods at 129 steps. The event was not stafety significant and only
                          resulted in an equipment problem not being identified as soon as it could have
                          been. The general superintendent of operations was reevaluating the reactor
                          startup procedure and c.onsidering adding hold points to refer to the procedure
                          and conduct the various verifications rather than expect the RO to remember the
                          entire C1B Procedure. This non-repetitive, licensee luentified and corrected
                          violation is being treated as a Non-C;ted Violation, consistent with Section Vll.B.1
                          of the NRC Enforcement Policy (50 282/97023-01(DRP)).
            c.  Conclusions
                Unit i startup operations were generally conducted well with no significant problems.
                Procedures were followed and operators remained attentive to plant indications during
                plant mode changes. The ROD AT BOTTOM annunciator cleared verification
                requirement in Appendix C1B, Step 5.5.8.C, which should have been performed at
                approximately 20 steps on control bank A, was not performed until 129 steps, primarily
                because the procedure was not required to be in-hand during the actual startup evolution.
                This error emphasized the need for improvements in pocedure organization and further
                evaluation of procedure use expectations. The licensee was evaluating possible
                improvements.
          01.3 Unit 1 Power Reduction for Turbine Balancina and Retum to Fu!! Power Operatim
            a,    Inspection Scope f71707)
                The inspectors observed significant portions of the Unit 1 power reduction frura
                  100 percent to approximately 5 percent power, the turbine balancing evolution, and the
                  subsequent retum to 100 percent power operation conducted from January 9 to
                  January 11,1998,
            b.    Observations and Findinas
*
                  The power reduction from 100 percent to 5 percent power was conducted very well. The
                  inspectors observed excellent communications between all operating crew personnel,
                                                              8
    .  .      .
                                  ..            .
                                                    .
                                                                        _ - - - .
 
  ._. -. . . .                    _ - - - -            .  -- -              -    - - -__        - . . - _ - - . . - - - -.-
            .  ... o
: ,. .                                                                                                                        ;
                                                                                                                              1
;
                          both inside and outside the control room. The RO and LPE&RO malntained good control
                          of the plant and awareness of plant parameters, and frequently kept each other informed
                          of changes in these parameters as indicated on the control room panels. The LPE&RO
                          executed the require!.ients specified in the Unit 1 power reduction (101.4, " Unit 1 Power
                          Operation,' Revision 15) and shutdown (1C1.3, * Unit i Shutdown,' ."evision 38)
'
                          procedures without error and effectively planned ahead to begin mwntenance activities at
                          the earliest appropriate opportunity The SS maintained overall cognizance of the power
                          reduction evolution, appropriately observing selected actions of the operators.
4
                          A second operating crew was observed performing the turbine stariup after the balanc!ng.
                          Again, excellent communications and plant control were exhibited. Procedures were
                          property implemented and the activities were closely supervised. The LPE&RO
                          controlling the turbine kept the RO closely informed of any changes that could affect the
                          reactor. The LPE&RO was also observed correcting a system engineer when the
                          engineer failed to use three way communications over the radio,
                      c.  Conclusions
'
                          The Unit 1 power reduction, turbine balancing, and retum to full power were conducted
                          well. Significant improvements were noted c.ompared to the control room etservations
                          desensed in Section 01.1 of this report.
                    02    Operational Status of Facilities and Equipment
                    O2.1 Engineered Safety System Walkdown
                      a.  Inspection Scope (71707)
                          The inspectors performed a walkdown of the Unit 2 containment spray and caustic
-
                          addition systems as part of the monthly inspections of the Unit 2 engineered safety
                          syt,tems. Included in this inspection was a review of USA't, Section 6.4, * Containment
                          Vessel Internal Spray," Revision 14, and the following diagram::
                          *          NF 39252, " Flow Diagram Unit 1 & 2 Caustic Addition System," Revision N;
                          e          NF 39824, " Containment Intemal Spray System Units 1 & 2," Revision B;
                          e            NF 39237, * Flow Diagram, Containment Internal Spray System," Revision AB; and
-                          *          NF 393331, * Reactor Safety injection ar'i Containment Spray Piping-Unit 2,"
                                        Revision R.
                      b.  Observations and Findinas
                          The inspectors noted that the valves in the main system f'tw paths were in the correct
                          position, components were properly labeled, locking devices were present and properly
                          installed, and that power supplies and breakers were properly aligned to support intended
                          system operation. No discrepancies were noted when comparing the system
                          components and layout with the system dese.riptions in the USAR. The material condition
                          of the systems was generally good, with the following exceptions:
                                                                        9
 
    _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _                                    _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _                          -____
                                    . . ,
                ,
                                                    e
  ..e
                                                            e      heavy buildup of sodium hydroxide crystals on the valve siem and packing gland
                                                                    area of valve 2 CA 20 5; and
                                                            e      light buildup of cheinical residue on the packing glands of valves MV 32110,
                                                                    MV 32111, CS 251, CS 40, CS 42, and 2 CA 19-4.
                                                            The discrepancies did not affect system operability.
.
                                                        c.  Conclus[gnt
                                                            During system walkdowns, the inspectors found the Unit 2 containment spray and caustic
                                                            addition systems lined up and ready for safeguards operation. No significant material
                                                            discrepancies or system deficiencies were identified that would prevent either system
                                                            from performing its intended function.
                                                      08    Miscellaneous Operations issues (92700,92901)
                                                      08.1  (Closed) Licensee Event Report 50 306/97005 (2 97-05): Sudden Pressure Lockout of
                                                            No.10 Transformer Resulting in Auto Load Rejection /Restcration on Safety Related Bus.
                                                            This event was previously discussed in Inspection Report No. 50>282/97021(DRP);
                                                            50-306/97021(DRP), Section 01.3. The Licensee Event Report (LER) contained a
                                                            detailed description of the event, investigation, and restoration sequence. Despite an
                                                            extensive investigation, no cause for the actuation was detem,ined, but the sudden
                                                            pressure relay was replaced as a precautionary measure. No additional corrective
                                                            actions were initiated since the causa of the actuation was not known.
                                                                                                                            II. Maintenance
                                                      M1    Conduct of Maintenance
                                                      M1,1 General Cornments
                                                        a.  inspection Scope (61726. 62707)
                                                            The inspectors observed all or major portions of the following maintenance and
                                                            surveillance activities. Included in the inspection was a review of the surveillance
                                                            procedures (SPs) and work orders (WOs) listed as well as the appropriate Updatad
'
                                                            Safety Analysis Report (USAR) sections regarding the activi'ies. The inspectors verified
                                                            that the surveillance procedures reviewed met the requirements of the TSs.
                                                              e      SP 1018A        Rod Position Indication Cold Calibration, Revision 6
                                                              *      SP 1070          Reactor Coolant System Integrity Test, Revision 26 (400 pounds
                                                                                      per square inch inspection portion only)
                                                              *      SP 1083          Unit 1 integrated Safety injection Test with a Simulated Loss of
                                                                                      Offsite Power, Revision 24
                                                                                                                                    10
                                                                -          -                                                              . - _ -
 
                                                                                          __
    ,
      .. . c
. . .
                *        SP 1089        Residual Heat Removal Pumps and Suction Valves from the
                                        Refueling Water Storage Tank, Revision 46
                *        SP 1194        Cardox [Carben Dioxide) System Test, Revision 8
                *        SP1231        121 Catalytic Hydrogen Recombiner Gas Analyzer Monthly
                                        Functional and Calibration Test, Revision 11
                *        LF 1301        11 Turbine Driven Auxiliary Feedwater Pump Auto Start and
                                        Function Testing Revision 8
                *        SP 1750        Post Outage Containment Closeout inspection, Revision 14
                *        SP 2548        Analog Reactor Control System Calibration, Revision b
                o        WO 9601293 Motor Operated Valve 32077 Excessive Packing Leakage
                *        WO 9706505 Fuel Oil Storage Tank Tightness Tests
                *        WO 9711674 Modify Tubing for 122 Control Room Chiller
                *        WO 9711686 Modify Tubing for 121 Control Room Chiller
                  *      WO 9712636 Turbine Torsional Test
                  *      WO 9713386 Annunciator 47013-0407 Doesn't Clear
                  *        WO 9713389 G3 and C7 Rod Bottom Bistables Will Not Clear
                  *        WO 9713491 Balance Unit 1 Turbine
                  *        WO 9715204 Possible Foreign Object Noise on DMIMS (Digital Metal Impact
                                          Monitoring System) Channels 750/751
                  *        WO 9715218 Monitor DMIMS Channels 750 and 751 for Noise
                  *        WO 9800004 High Average Coolant Temperature Compensator Module Spiking
              b.  Observationi md Findinas
                  For all of the work observed, procedures were properly used and followed except for one
                  activity discussed in Section M1.2 of this report. Maintenance personnelwere
                  experienced and knowledgeable of their tasks. The inspectors observed frequent
                  monitoring of work by system engineers. Noteworthy comments on specific work
                  activities are discussed below.
                  *      For SP 1070, the inspectors accompanied two operators and a radiation
                          protection technician on an inspection for indications of reactor coolant system
                          leakage in the Unit 1 containment. Two prejob briefings were held: one involved
                          the operators and the shift supervisor to discuss the technical aspects of the task,
                                                            11
                                                                                                                l
                                                                                                                i
 
                                                                                                .. __
        ....
  . . .
!
              and the other involved the operators and a radiation protection superviser to
              discuss the radiation protection aspects. The two operators split the areas to be
              inspected and carefully planned their routes with the ascistance of the radiation
              protection technician to minimize radiation exposure. The inspections were
              conducted expeditiously, but thoroughly, No indications of leakage were
              identified. The operators displayed a suitable approciation for maintaining their
              radiation dose as low as reasonably achievable wt'lle still property conducting the
              inspections.
            * For SP 1089, the operators identified a proceduru enhancement during the
              conduct of the test. The test required the operators to record differential pressure
              (D/P) in the residual heat removal pump mini-flow recirculation line. Since the D/P
              gauge exhibited fluctuations around the actual reading, the procedure allowed the
                option of throttling the instrument root valves to dampen the fluctuations. The
                operators properly followed the procedure and throttled the root valves, but
                performance of that step was quite difficult. The root valves were located in the
                next room from the gauge,in a contaminated area, approximately 10 feet off the
                floor. The gauge was not visible from the va'ves. Close coordination of two
                operators was necessary to complete the task. The operators identif5d that it
                would have been much easier to throttle the instrument isolation valves on the
                manifold just under the gauge. The chift supervisor initiated action to get the
                procedure revised.
                Just before throttling the instrument root valves, the local operators questioned
                whether there was any other indication or actuation circuitry off the same
                instrument lines. They were concer,.ed that they might accidently isolate the line
                by overthrottling and cause some kind of problem in other instrumentation. The
                operators stopped and resolved their concem with the shift supervisor before
                proceeding. The operators displayed a conservative operating philosophy and
                questioning attitude during the test,
            e  On December 15,1997, the inspectors observed activities govemed by
                WO 9712636, * Unit 1 Turbine Torsional Testing,'' required after the replacement
                of the turbine generator low pressure rotors. The test measured the torsional
                natural frequencies and response levels of the turbine generator shaft system.
                Step 8.1.8.f of the turbine torsional test procedure instructed the contrei room
                operator to close generator output breaker 6 H 17. The control room operator
                displayed a questioning attitude by asking the system engineer if it was necessary
                to place the synchroscope selector switch in the BKR 17 position prior to closing
                B H 17. The system engineer responded that this was not necessary since the
                synchro-check relay had been bypassed. When the control room operator
                attempted to close 8-H 17, in accordance with Step 8.1.8.f of the WO, the breaker
                did not close. Further review of the electrical schematics with the system
                engineer revesk ; that, although a jumper had been installed to bypass the
'
                synchro-check relay in the sivitchyard, the procedure failed to recognize that the
                synchroscope selector switch in the control room needed to be placed in the
                BKR 17 position to makeup auxiliary contacts necessary to close 8-H-17. The
                                                                  12
                                                  _ _ _ _ _ _ - _
 
      . 4 .
  .. .
                        WO was changed, the synchroscope selector switch was placed in the BKR 17
                          position, and 6 H 17 was closed with its control switch. The Unit 1 turbine
                          torsional test was completed without further problems.
                          The turbine torsional test procedure was not appropriatt for the circumstances
                          because it did not require placing the synchroscope selector switch in the BKR 17
                          position prior to closing 8 H 17. This was a violation of 10 CFR Part 50,
                          Appendix B, Criterion V, which required that activities affecting quality be
                          prescribed by docume:ited instructions, procedures, or drawings, of a type
                          appropriate to the circumstances. However, the turbine torsional test procedure
                          problem was not safety significant becauso it simply resulted in the inability to
                          complete the test untilit was corrected. This non repetitive, licensee-identified
                          and corrected violation is being treated as a Non Cited Violation, consistent with
                          Section Vil.B.1 of the NRC Enforcement Policy (50-282/97023-02(DRP)).
                  e        For WOs 9715204 and 9715218, the licrasee noted, soon after starting Unit i bp
                          from a refueling cutage, metallic noise indications on detectors located on the
                          reactor vessel. The licensee made recordings of the noise and sent them to
                          vendors for analysis and made plans to try to identify the source. However, during
                          the power reduction on Januery 9, the noise stopped as soon as operators started
                          to insert control bank D control rods. The licansee wes unable to reestablish the
                          noise by withdrawing control bank D rods while at reduced power, but the noise
                          recurred when reactor power was at about 95 percent and control bank D rods
                          were retumed to about 205 steps out. Thus, the licensee believed the noise was
                          associated with at least one of the control rods and did not represent en
                          immediate safety concern. At the end of the inspection period, the licensee was
                          considering further actions to identify the exact cause of the noise.
            c.  Conclusions
                  Operators involved with maintenance and surveillance activities displayed a good
                  questioning attitude and appreciation of mdiation dose control. A procedure inadequacy
                  was identified where circuit logle was :,st analyzed thoroughly enough during
                  devolopment of a work order procodure. System engineers frequently mor:ltored ongoing
                  work and were generally responsive to maintenance staff concems: ho...for, in one
                  instance, during the turbine torsional testing, a system engineer did not provide adequate
                  support.
            M1.2 Low Power Physics Testina
            a.  Inspection Scope (71711)
                  The inspectors observed the conduct of various maintenance activities for refueling
                  startup physics testing on Unit 1. The following maintenance procedures and documents
                  were reviewed as part of this inspecticn:
                  *          D32, " Temperature Coefficient Measurement at Hot Zero Power," Revision 6;
                  *            D34, " Boron Endpoint Measurement,' Revision 6;
                                                              13
A
                          . _ _ _ _ _ _ _ _ _ _ _ _
 
                                                                    -______--
    o
.. .
          *          D30, * Post Refueling Startup Testing," Revision 27; and
          *          ANSI /ANS (American National Standards Institute, Inc./American Nuclear Society)
                    19.6.1 1985, * Reload Startup Physics Tests for Pressurized Water
                    Reactors."
      b. Observations and Findinas
          The performance of maintenance activities associated with Procedure D34 required a
          significant number of control rod position manipulations. The reactor operator (RO)
          exercised good control over reactivity during the selected individual manipulations of all
          the shutdown and control bank control rods. Good coordination was observed between
          the reactor operator and the nuclear engineer assisting with the procedure. Good
          supervisory oversight was provided by the shift supervisor in that he provided a second
          verification that the correct rod bank was selected prior to control rod movement.
          The inspectors also observed test activities associated with Maintenance Procedure D32.
          These activities were required to be performed twice during low power physics testing:
          once with all the control rods withdrawn and once with all rods withdrawn except for the      [
          control bank A, which was fully inserted. The purpose of the procedure was to determine
          the isothermal temperature coefficient (lTC) at an established condition below the point of
          adding heat and to verify that it was less that 5 percent millirho per degree Fahrenheit, as
          required by TS 3.1.F,1.
          The actual performance of the test, after initial plant conditions had been ostablished,
          was accomplished by performing a reactor coolant system (RCS) cooldown followed by a
          heatup. During each trsasient, a plot of reactivity versus temperature was obtained
          during which time boron concentration and control rod position were kept essentially
          constant. The magnitude of the cooldown and heatup, as required by Steps 7.2 and 7.3
          of Maintenance Procedure D32, was approximately 5 degrees Fahrenheit ('F). More
          specific guid1nce was contained in ANSI /ANS 19.6.1 1985 which was listed as a
          reference for D32. Standard ANSI /ANS 19.6.1 1985 stated, as part of the test method to
          determine the ITC, that reactivity and temperature should be continuously recorded while
          RCS temperature is increased (decreased) by 3-10 *F.
          During the second test per D32, the inspectors identified that an RCS cooldown of 2.4 'F
          and an RCS heatup of 1.2 'F were used to determine the value for ITC. When the
          inspectors questioned the nuclear engineer conducting the test about the procedural
          requirement for an approximate 5 'F cooldown (heatup) while obtaining ITC data, i'.ie
          nuclear engineer said that a sufficient RCS temperature change had been performed and
            that the data was good enough to calculate ITC. The inspectors also discovered that the
            first time the ITC test was performed with all control rods out, a 2.3 'F cooldown and
            a 1.8 *F heatup were used. These discropancies were brought to the attention of the
            superintendent of nuclear engineering. He agreed that Maintenance Procedure D32 was
            not followed as written. He also said that because it was difficult to maintain a steady
            cooldown (heatup) over a range in excess of about 3 'F, that the cooldown (heatup)
            portions of the test had usually been performed with less than an approximately 5 *F
            temperature change for many years.
                                                      14
                                                                ___
 
    , , ,
, .
                The inspectors detormined that sufficient temperature changes were accomplished to
                obtain adequate data for calculation of the ITC, so this violation was of low safety
                significance. However, failure to perform the ITC test per Procedure D32 as written, or
                revise the procedure, despite numerous opportu,nties, demonstrated a lack of
                appreciation for procedural requirerrents. There have been several cited violations for
                failure to follow procedures documented in previous inspection reports. In addition, the
                importance of procedure adherence was one of the topics in two recent management
                meetings Corrective actions Mr the previous violations should have prevented this
                procedure noncompliance, in addition, this violation was identified by the NRC.
                Therefore, the event did not meet the criteria for discretion in the NRC Enforcement
                Policy.
                The failure to perform maintenance activities associated with Procedure D32 as written
                was a violation of TS 6.5.A.4, which required that the licensee prepare and follow detailed
                written procedures that control surveillance and testing requirements that could have an
                effect on nuclear safety (50-282/97023-03(DRP)).
          c.  Conclusion
                The inspectors concluded that both the operators controlling the plant and the nuclear
                engineers coordinating the performance of maintenance activities associated with
                Procedure D32 Jid not follow the written instructions provided in the procedure pertain'ng
                to the magnitude of cooldown (heatup) required forITC de armination. Even though the
                engineers knew the proceduie requirements, they chose to do what had worked in the
                past, instead of evaluating the way they were performing the test and changing the test to
                reflect the actual test practice. This demonstrated a lack of appreciation for strict
                compliance with procedural requirements.
          M3    Maintenance P.-ocedures and Documentation
          M3.1 Coo:ina Water System Walkdown and Suweillance Procedure Review
          'i . Inspection Scope (71707. 62707)
                The inspectors conducted a walkdown of the Unit 1 and Unit 2 cooling water systems,
                included in the inspection was a review of USAR, Section 10.4, * Plant Cooling System,"
                Surveillance Procedures SP 1168.8, * Cooling Water System Operating Pressure Test,"
                Revision 9, and SP 1168.8A, * Cooling Water System Auxiliary Operating Pressure Test,"
                Revision 0, and a detailed review of the following American Society of Mechanical
                Engineers (ASME) Code drawings:
                *      NF 39819-1,'Cooiing Water ASME Code Classification Screenhouse Unit 1,"
                        Revision B;
                *      NF 39819-2, * Cooling Water ASME Code Classification - Turbine Building Unit 1,"
                        Revision B;
                *      NF 39819 3, * Cooling Water ASME Code Classification - Auxiliary Building Unit 1,"
                        Revision D;
                                                          15
 
  - , _ - - -_-___ _,. - _ . - -                            .  - -        ..      .                - . - - - - . . . - -        - - - . _
              ,
                    . .. c
      .< ,
                                    e      NF 39819-4," Cooling Water ASME Code Classification Containment Unit 1,
                                            Revision C;
                                    e      NF 39841 1, * Cooling Water ASME Code Classification Turbine Building Unit 2,*
                                            Revision A;
                                    e      NF 398412," Cooling Water ASME Code Classification Auxiliary Building Unit 2,*
                                            Revision E;
                                    e      NF 398413,' Cooling Water ASME Code Classification Containment Unit 2,*
                                            Revision C;
                                                                                                                                              i
                                    e      NF 39822, * Prairie Island Nuclear Generating P! ant Fuel and Diesel Oil System -                  I
!
                                            Units 1 & 2 ASME Code Classification Sheet 17," Revision A; and                                    l
                                                                                                                                              )
                                    *      NF 39833," Lab Service Area and Chilled Water Safeguards Systems -                                1
                                            Unt'.s 1 & 2 ASME Code Classification - Sheet 26," Revision D.                                    I
                                                                                                                                              '
                                b. Observations and Findinas
                                    The inspectors observed the material condition of the Unit 1 and Unit 2 cooling water
                                    systems and did not identity any significant discrepancies. All equipment and systems
                                    matched the description found in USAR Section 10.4. However, surveillance Procedures
,                                  SP 1168.8 and SP 1168.8A contained 21 discrepancies.
                                    SP 1168.8 and SP 1168.8A were required by TS 4.2 and the licensee's ASME Code
                                    Section XI Insery!ce Inspection and Testing Program. Each surveillance was performed
                                    at leasi once every 3% years. The surveillance directed that personnellook for evidence
                                    of component leakage, structural stress, and corrosion, and that they inspect hangers
                                    and restraints to detect any loss of support capability, missing or loose bolts, corrosion,
                                    ed other problems.
                                    SP 1168.8 contained the following procedural discrepancies:
                                    e        a 12" diameter section of the cooling water supply line to the 22 component
                                            cooling water heat exchanger was not required to be inspected by SP 1168.8;
                                    e        a 24" diameter section of the Unit 1 cooling water retum header located just
                                            upstream of the auxiliary building / turbine building wall penetration was not
                                            required to be inspected by SP 1168.8;
                                    e      two instances where SP 1168.8 listed valves not shown on the ASME Code
                                            drawings;
                                    e      three instances where the valve designations on the ASME Code drawings did not
                                            match the valve numbers included in SP 1168.8;
                                    e      one instance where ASME Code drawing NF 39819-3, Revision B, showed two 1"
                                            diameter cooling water is otation valves in the same line when only one existed in
                                            the plant; and
                                                                                16
                                                                                                                          - , ..          . -
 
      4
        o
  ,. .
                                                                                                                                  O
I                  *        une instance where two separate steps called for inspecting the same section of
                            the 24* loop A cooling water header.
          .
                  SP 1168.8A contained 12 instances where valves named in the procedure as requiring
                  inspection were shown but not labeled on the referenced ASME Code drawings.
                  The inspectors noted that both SP 1168.8 and SP 1168.8A contained the same                                    *
                  precaution and limitations section S'.ep 3.2 stating,"The individual sign off steps are
                  intended as a guide; use the Code drawings and/or isometrics to verify alllines are
                  inspected." In most cases, using the ASME Code drawings and carefully tracking each
                  section of line inspected would preclude the inspector from missing portions of the Unit 1
                  and 1.' nit 2 cookng water system not described in SP 1168.8 or SP 1168.8A. In at least
                  two cases, however, SP 1168.8 or SP 1168.8A included the inspection of valves not
                  described on the ASME Code drawings. By usli '; the surveillance procedures, Code
                  drawings, and Isometric drawings as directed, individuals cc a orrectly perform the
                  inspection. Therefore, the procedures were not considered ... adequate.
                  The inspectors discussed the above findings with the system engineer prior to the exit
                    interview. The system engineer took prompt action to correct SP 1168.8, SP 1168.8A,
                    and the ASME Code drawings for the inspector identified discrepancies note.1 above,
              c.  Conclusions
                    Several procedural deficiencies were identified in SP 1168.8 and SP 1168.8A. None of
                    the deficiencies were safety significant because the surveillance also required the use of
                    drawings to confirm that all sections of piping were inspected. However, the deficiencies
                    made the inspection task more difficult.
                    The cooling water system had recolved a great deal of review over the previous three
                    years in a licensee self assessment and NRC inspections. The inspectors were
                    concerned that a system that had received so much recent attention could still have
                    procedures containing so many errors. The inspectors considered SP 1168.8 and
                    SP 1168.8A reflective of the need to further improve pmcedures.
                    The licensee recently completed a pilot program to review a sampling of surveillance
                    procedures for accuracy and compilance with the writer's guide. Errors were reportedly
                    found in each of the surveillance procedures reviewed. The licensee was making plans
                    to extend the scope of the procedure review in light of those findings.
            M3.2 Procedural Weaknesses Identified Durina Auxiiiarv Feedwater Pumn Testina
                                                                                                                                  ,
              a.    ingoection Scope (61726)
                    The inspectors attended the prejob brief and observed the performance of testing per
                    SP 1301, "11 Turbine Driven Auxiliary Feedwater Pump A a Start and Function Testing "
                    Revision 8. The inspectors reviewed SP 1301 for procedural adequacy and compliance
                    with TS 4.8.A.8 and Table 4.1-1C, items 26 and 27, and USAR Section 11.9.
                                                              17
                                                                                                                                    !
                                                                                                                                    l
                              _.                                                                  -            -_-------_--.---.---J
 
      ,,
11 '
            b.      Observations and Findinns
                    The projob brief for SP 1301 was thorough and discussed communications, impacts of
                    ongoing Unit 2 electrical plant shifts with the surveillance, worker responsibilities, and the
                    procedural steps involved in each of the five functional areas being tested. During the
                    performance of the surveillance; however, three typographical errors were noted. Two
                    errors were Identified by the inspectors observing the evolution and one by the control
                    room operator supervising the surveillance. These errors are described below.
      _
                    Stop 7.10.28 directed instrument and control personnel to positi,:a three switches in the
                      1 ARP5 Reactor Protection Logic Test Cabinet.1 he three switches in this cabinet were -
                    labeled 81, S2, and S3. Step 7.10.28, however, contained a typographical error and-
                    referred to the switches as S1,82, and S2. The error was so obvious that the
                    technicians performing the surveillance had no problems. The inspectors brought the
                      error to the attention of the operator who notified the control room.
                      Stop 7.11.5 verified that the AMSAC [ Anticipated Transient Without Scram Mitigation
                      System Actuation Circuit] INACTIVE annunciator in the control room was ON. The step,
                      however, contained a typographical errnt and identified the annunciator location as
                      47074 0606 when the correct location was 47014-0606. The noun name description for
                      the annunciator was correct in the procedure so it did not cause a performance problem.
                      The operator supervising the surveillance ir, the control room identified this error,
                      in Step 7.15.6, an operator was directed to push the 11 turbine-driven auxiliary feedwater
                      pump local stop push button Step 7.15.6 con:sined a typographical error and r.pecified
                      depressing the time delay reset push button PS 5101801 instead of the actuallocal stop
                      push button, PB-5101803. The noun name description in the procedure was correct so
                      the operator performed the correct action, even though the push button number was
                      incorrect. The inspectors observing the surveillance noticed this error and brought it to
                      the attention of the operator. The operator informed the control room.
                      The control room operator supervising performance of SP 1301 submitted procedure
                      deviation requests to correct the errors noted above following completion of the
                      surveillance.
                  c.  Corclusions
                      The three typographical errors identified in SP 1301 had no safety significance and did
                      r,ot prevent satisfactory performance of the surveillance. The inspectors were concemed,
                      however, that two of the three errors were NRC-identified and not noticed by the five
                      licensee personnel (one electrician, one instrument and control technician, two outplant
                      operators, and one control room operator) actually conducting the surveillance, in the
                      case of the two NRC-identified errors, the test performers did not use proper self-
                    = checkin2 techniques. They did not adequately check the switch numbers listed in the
                      procedure against the actval component label before completing the action.
                      SP 1301 was reviewed by the Operations Committee on Coptember 17,1997, and
                      approved by the superintendent mechanical systems on November 29,1997. This was
                      after the licensee placed a renewed emphasis on procedural adequacy and compliance.
                                                          '
                    - The three errors identified li. ::P 1301 highlighted the need for continued efforts in this
                                                                            18
    .
        .. . .. .              ..
                                                            .
                                            _ _ _ _ _ _ .    _ _ _ _ _ _ _
 
                                                                                    . _ _ _ _ _ _ _ _ _ _ _ - -_-_-
  _
        4 ,
    .: .
                    area. As noted in Section M3.1 of this report, the licensee was evaluating an expanded
                    surveillance review program.
            M8    Miscellaneous Maintenance Activities (92700,92902)
)
'
            M8.1 (Closed) Inspection Followup Item (IFI) 50-282/97005 03fDRP): 50 306/97005-03(DRP):
                    Reactor Coolant System Vent and Containment Boundary Control During Integrated
                    Leakage Rate Test. This issue was previously discussed in inspection Report
                    No. 50-282/97005(DRP); 50-306/97005(DRP), Section M1.1. It involved a licensee-
                    identified procedure error in Unit 2 surveillance Procedure SP 2071.4, * integrated
                    Leakage Rate Test Prerequisites to the Containment Vesse! Integrated Leakage Rata
                    Test," Revision B. The test procedure was written in such a way that the reactor coolant
                    system could be vented before containmg.t integrity was established, which would have
                    been a violation of TSs. The inspectors verified that the Unit 1 test, SP 1071.4,
                    " Prerequisites to the Containment Vessel Integrated Leakage Rate Test," Revision 6, had
'
                    been revised to eliminate the problem before it was used. The inspectors also verified
                    that the system engineer had submitted a procedure revision form to ensure that the
                    Unit 2 procedure would be revised before its next use. The next expected Unit 2
                    integrated leakage rate test was planned to be conducted in the year 2006.
                                                                Ill. Enoineerina
            E2      Engineering Support of Facilities and Equipment
            E2.1    Review of Updated Safety Analysis Report (USAR) Comntitments (37551. 92903J
                    While performing the inspections discussed in this report, the inspectors reviewed the
                    applicablo portions of the USAR that related to the areas inspected and used the USAR
                    as an engineering / technical support basis document. The inspectors compared plant
                    practices, procedurcs, and/or parameters to the USAR descriptions as discussed in each
                    section. The inspectors verified that the USAR wording was consistent with the observed
                    plant practices, procedures, and parameters. No discrepancies were noted.
            E2.2    General Comments (37551)
                    Throughout the inspection period, 'Se inspectors noted frequent involvement by system
                    engineers in all aspects of plant operations, refueling, maintenance, and surveillance
                    activities. The engineers rapidly investigated any operational abnormalities, took an
                    active role in maintenance and troubleshooting activities, and closely followed all
                    surveillance testing on their systems.
                                                                          19
                                        - - __        _ _ _ _ _
 
    .
      ..
  .,
i        E3  Engineering Procedures and Documentation
          E3.1 ftilure to Test the Auto Start Feature of the Control Room Ventilation System Air
              Handlers Due to Procedure Deficiency
          a.  Inspection Scope (g2700)
              On December 7,1997, as part of an investigation in response to NRC Generic
              Letter 96 01, " Testing of Safety related Logic Circuits," the licensee discovered that the
              automatic start of the 121 and 122 control room air handlers upon a st:rt of the
              associated 121 and 122 control room cleanup fans, which was required to be tested by
              TSs, had not been tested. The inspectors reviewed the circumstances and corrective
              actions for the finding,
          b.  Observations and Findinos
                Technical Specification 4.14.A.2 required, in part, that once per operating cycle or once
                every 18 months, whichever occurs first, the automatic initiation of the control room
                special ventilation system be demonstrated with a simulated high radiation or safety
                injection signal. A high radiation or safety injection signalis designed to start the control
                room cleanup fans and operate various dampers. Starting of the clesnop fans will
                subsequently result In the start of the main air handler fans. Without the air handler fans
                running, the cleanup fans would be ineffective in performing the function of reducing
                airborne radioactivity for control room operators.
                The licensee discovered that the surveillance had normally been per4ormed with the air
                handler fans already running in order to test the isolation function of the outside air
                dampers. Thus, one of the automatic features had not been tested.
                As discussed in the Licensee Event Report (LER 19718) for the finding, shortly before
                the time of discovery, the auto start feature of both trains of control room ventilation had
                coincidently been tested as part of a pre-operational test for a modification of the related
                power supplies. The auto start features functioned normally during those tests. While
                this indicated that the air handler fans' automatic start feature was functional, the failure
                to test it as part of a formal surveillance test reflected a programmatic weakness in tha
                surveillance testing program. The LER stated that the survelliance proceduret would ue
                revised prior to the next scheduled test to require testing of the auto start feature. The
                LER will rema.in open pending completion of the revisions (LER 50-282/97018;
                50-306/97018).
                The failure to perform a surveillance test of the auto start feature of the control room alt
                handlers was a violation of TS 4,14.A.2. This non-repetitive, licensee-identified and
                corrected violation is being treated as a Non-Cited Violation, consistent with
                Section Vll.B.1 of the NRC Enforcement Policy (50-282/97023-04(DRP);
                50-306/97023-04(DRP)).
            c.  Conclusions
                  This finding, and the ones discussed in Sections EB.1 and F2.1 of this report indicated
                  that the licensee was conducting thorough design reviews in response to Genotic
                                                            20
                                                            - - _ - -
 
    .*e
      .
  ..
                Letter 96 01 and other concems. Licensee employees demonstrated a wil:ingness to
                identify old design discrepancies and compliance problems and the licensee rapidly
                resolved those issues.
        E8      t,11scellaneous Engineering issues (92700,92903)
        E8.1    (Clesed) LER 50 282/97015: 50-306/97015 fi 9715h Both Trains of Control Room
                Special Ventilation System Simu'taneously Inoperable. This LER discussed an issue,
                identified on November 'J,1997, in which the licensee determined that routine
                performance of a monthly surveillance on steam exclusion dampers had resulted in both
                trains of control room sponial ventilation being inoperable because outside air dempers,
                opened for the surveillance, would not have automatically closed on actuation of the
                  system on high radiation or safety injection. This could have resulted in the control room
                operators' dose being higher than General Design Criterion 19 limits.
                The cause of the event was the failure to properly review the control room ventilation
                cystem logic and design requirements when the steam exclusion damper surveillance
                was devcloped. The Feensee identified the issue as part of the development of revised
                  main steamline break control room dose calculations. As discussed in the LER, the
                  condition existed for only a few minutes each month and the operators would have had
                  ample indications and controls available to identify the problem and close the dampers if
                  an accident had occurred. The inspectors verified that all of the corrective actinns
                  discussed in the LER had been completed.
                  Monthly performance of the steam exclusion damper surveillance resulted in both trains
                  of control room special ventilation being inoperable, contrary to the requirements of
                  TS 3.13.A.1 which required that both trains be operable at all times. Although the TS
                  was violated, the associated action rwquirement to initiata within 1 hour the action
                  necessary to place both units in hot shutdown, and be in at least hot shutdown within the
                  next 6 hours and in cold shutdown within the following 30 hours and terminats core
                  alterations / fuel handling operations within 2 hours, was probably never exceeded.
                  This non repetitive, licensee identified and corrected violation is being treat?d as a
                  Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy
                  (50 282/97023-05(DRP); 50 306/97023 05(DRP)).
                                                    IV. Plant Suooort
          R1      Radiological Protection and Chemistry Controls (71750)
          During normal resident inspection activities, routins observations were conducted in the areas of
          radiological protection and chemistry controls using inspection Procedure 71750. No
          discrepancies were noted. The inspectors noted good involvement of radiation protection
          personnel in job planning and execution in order to maintain doses as low as reasonably
          achievable.
                                                            21
<
                                                .
 
                                                                        -                _          -_
      ..e e
  .- .
            P1    Conduct of Emergency Preparednees Activities (71750)
            During normal resident inspection activities, routine observations were conducted in the area of
            emergency preparedness using Inspection Procedure 71750. No discrepancies were noteo.
            81      Conduct of Security and Safeguards Activities (71760)
            During normal resident inspection activities, routine observations were conducted in the areas of
            security and safeguards activities using Inspection Procedure 71750. No discrepancies were
            noted.
            F2      Status of Fire Protection Facilities and Equipment
            F2.1    Separation of Pressurizer Level Indication Channels Not in Comollance with
                    10 CFR Part 50. Accendix R. Section Ill.G.2
              a.    laspection Smoe (92700)
                    On December 6,1997, the licensee reported to the NRC in accordance with
                      10 CFR 50.72 that the plant was in a condition outside of the design basis because the
                    licensee had discovered that the pressurizer level channel cables in Unit 1 containment
                    were not separated as required by 10 CFR Part 50, Appendix R. The inspectors              r
                    reviewed the circumstances and corrective actions for the finding,
              b.    Qbtervations and Findinas
                      The issue was first identified during a walkdown of the containment on
                      November 14,1997, to support an Appendix R Safe Shutdown Analysis revision. The
                      licensee noted that the pressurizer level detectors were not located as shown on plant
                      layout drawings and that the pressurizer level channels old not have adequste separation
                      of the detectors and the associated cables to meet the requirements of Appendix R. The
                      drawings indicated adequate separation but the as Lailt configuration did not match the
                      drawings.
                      At first the licensee believed that an exemption from the NRC might have been grants d
                      for the existing installation. As discussed in the associated Licensee Event Report
                      (LER 1-97-17), the documentation regarding various Appendix R exemptions was
                      somewhat confusing and incomplete. By December 6, the licensee determined that an
                      appropriate exemption did not exist and the cabling would have to be modified to meet
                      Appendix R.
'
                      Within the next few days a modification was developed and installed to provide a
                      noncombustible radiant energy shield around one channel of the cabling to satisfy
                      Appendix R requirements. The inspectors walked down the installation and it appeared
                      that the energy shieH was adequate.
                      The licensee issued LER 19717 on January 2,1998. The inspectors reviewed the LER
                      and determined that it adequately discussed the beckground, safety significance, and
                      corrective actions for the subject Appendix R non compliance. The LER will remain open
                      pending completion of corrections to the plant layout drawings (LER 50-282/97017).
                                _                  __ __      . _ _
                                                                    22
          ..
                        .
                                        _ _ _ - _ - _ - _ _ _
 
          , ;e; .
    0*
*
                              Sedion lll.G.2 of App 9ndix R, of 10 CFR Part 50, requires, in part, that cables and
                              equipment and associated non safety circuits of redundant trains of equipnient necessary
4                            to achieve and maintain hot shutdown conditions in non-inerted containments, be
4
                              protected from potential fire Jamage. This could be achieved through separation of
a
'
                              more than 20 feet with no intervening combustibles or fire hazards or by having installed
                              fire detection tend automatic suppression, or through separation by noncombustible
j.                            radiant energy shields. The licensee 6dentified that the redundant pressurizer level
i                            detector cables in :he Unit 1 containment did not satisfy any of these conditions and thus
'!                            a violation of NRC requirements existed. As discussed in the LER, the condition had
j                            M!stively low safety significance and was promptly corrected when the noncon.pliance
j                            was confirmed. Thi6 non repetitive, licensee-identified and corrected violation is being--
                              treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement
i
                              Policy (50 282/97023-06(DRP)).
}                          c. Conclusions
f
;.                            The licensee's finding was the result of a proactive voluntary review of fire protection
i                            issues and revision of tire protection analysis. Prompt corrective actions were taken
3                            when the condition v as confirmed.
                                                                                                                                                                              '
l
                                                                                                                                                                              '
i
l-                                                                                  V Mananoment Meetinos
                                                                                                                                                                              ,
                      X1    Exit Meeting Summary
                      The inspectors presented the inspection retu;is to members of the licensee management at the
                      conclusion of the inspection on January 13,1998. The licensee acknowlectged the findings
4
                      presented. The inspectors asked the licensee whether any materials examined during the
                      inspection should be considered proprietary No proprietary information was identified.
$
:
.
                                                                                                                                                                              i
1
,
9
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                                                                                                      23
1
4
  y,-w m*t      ,~w%,,r*r-  w--#w.--*et'-***- 7 * 't F *' v'-r- ' -~ --T--wm 'w N    * '*v= '-v"w'=re-''***"*-**'''*-*---*'*-''"**'r'r''***' ' ' "'''*-''v*"* --
                                                                                                                                                                  #' '" ' " '
 
                                                                                        _ . - . = - - _ - - . - - - . .
    *
  ,..,=
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0-
                                  PARTIAL List OF PERSONS CONTACTED
        Licensee
        J. Sorontsn, Plant Manager
        K. Albrecht, General Superintendent, Engineering, Electrical / Instrumentation & Controls
        T. Amundson, General Superintendent, Engineering, Mechanical
        J. Goldsmith, General Superintendent, Engineering, Generation Services
        J. Hill, Manager, Quality Services
        G. Lon3rtz, General Superintendent, Plant Maintenance
        J. Maki, Outage Manager
        D. Schuelke, General Superintendent, Radiation Protection and Chsmistry
        T. Silverberg, General Superintendent, Plant Operations
        M. Sleigh, Superintendent, Security
                                _
                                                      24
                                                                                                                        ,
 
                  _                                    - - . .
    .
      . . .; .
  - .
1
!
l                                              lNSPECTION PROCEDURES USED
l
                lP 07551:      Engineering
                IP 61726:      Surveillance Observations
                IP 62707:      Maintenance observations
                IP 71707:      Plant Operations
                IP 71711:      Startup from Refueling
                IP 71750.      Plant Support Activities
                IP 92700:      Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor
                              Facilities
                IP 92901:      Follow up - Operations
                IP 92902:      Follow up - Maintenance
                IP 92903:      Follow up - Engineering
                                          ITEMS OPENED, CLOSED, AND DISCUSSED
                Qllen9A
                50-282/97023-01(DRP)        NCV      Failure to Perform Step in Reactor Startup Procedure
                                                      When Called For
                50-282/97023-02(DRP)        NCV      inadequate Procedure for Turtaine Torsional Testing
                50 282/97023-03(DRP)        VIO      Failure to Perform Two Steps in Reactor Physics Testing
                                                      Procedure as Written
                50-282/97023-04(CRP)        NCV      Failure to Test the / % Start Feature of the Control Room
                50 306/97023-04(DRP)                Ventilation System il- Handlers due to Procedure
                                                      Deficiency
                50-282/97023-05(DRP)        NCV      Both Trains of Control Room Special Ventilation System
                50-300'97023-05(DRP)                Simultaneously inoperable
                50 282/97023- 06(DRP)      NCV      Separation of Pressurizer Level Indication Channels Not in
                                                      Compliance with 10 CFR Part 50, Appendix R,
                                                      Section Ill.G.2
                50 282/97017                LER      Separation of Pressurizer LevelIndication Channels Not in
                                                      Compliance with 10 CFR Part 50, Appendix R,
                                                      Section Ill.G.2
                50-282/97018                LER      Failure to Test the Auto-Start Feature of the Control Room
                5B Gi97013                            Ventilebon System Air Handlers due to Procedure
                                                      Deficiency
                Closed
                50-282/97005-03(DRP)        IFl      Reactor Coolant System Vent and Containment Boundary
(.              50-306/97005-03(DRP)                  Control During Integrated Leakage Rate Test
                50-306/97005                LER      Sudden Pressure Lockout of No.10 Transformer Resulting
                                                      in Auto Load Rejection / Restoration on Safety-Related Bus
                                                                    25
 
    ~
(        .
    , ' ?. 0
  .
    .
            50-282/97015 LER Both Trains of Control Room Special Ventilation System
!            50-306/97015    Simultaneously Inoperable
1
                                          26
                                              _ _ _ _ _ - _ - _ _ -
 
                                                                        .. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _
,
    , . ',i %
  1*
                                            LIST OF ACRONYM 8 USED
              AMSAC    Anticipated Transient Without Scram Mitigation System Actuation Circuit
              ANSI /ANS Americcn National Standards Institute, Inc/American Nuclear Society
              ASME      American Society of Mechanical Engineers
              CFR      Code of Federal Regulations
              D/F      Differential Pressure
              DRP      Division of Reactor Projects
              DMIMS    Digital Metal impact Monitoring System
              'F        Degrees Fahrenheit
              IP        inspection Procedure
              ITC      isothermal Temperature Coefficient
              LER      Licensee Event 9eport
              LPE&RO    Lead Plant Equipment and Reactor Operator
              NRC      Nuclear Regulatory Commission
              PDR      Public Document Room
              RCS      Reactor Coolant System
              RO        Reactor Operator
              SP        Surveillance Procedure
              SS        Shift Supervisor
              SWI      Section Work Instruction
              TS        Technical Specification
              USAR      Updated Safety Analysis Report
              WO        Work Order
?:
                                                                      27
                                                    _ _ _ _ _ - _ _ -                                                  I
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