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{{Adams
#REDIRECT [[IR 05000354/1997010]]
| number = ML20199L532
| issue date = 02/03/1998
| title = Insp Rept 50-354/97-10 on 971116-980103.Violations Noted. Major Areas Inspected:Licensee Operations,Engineering,Maint & Plant Support
| author name =
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| addressee name =
| addressee affiliation =
| docket = 05000354
| license number =
| contact person =
| document report number = 50-354-97-10, NUDOCS 9802090222
| package number = ML20199L447
| document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 29
}}
See also: [[see also::IR 05000354/1997010]]
 
=Text=
{{#Wiki_filter:,.
  .
                            U. S. NUCLEAR REGULATORY COMMISSION
                                              REGION I
      Docket No:          50 354-
      License Nos:-      NPF 57
      Report No.          50 354/97-10
      Licensee:          Public Service Electric and Gas Company
      Facility: -        Hope Creek Nuclear Generating Station
l-
'
      Location:          P.O. Box 236
                          Hancocks Bridge, New Jersey 08038
      Dates:              November 16,1997 - January 3,1998
      Inspecters:        S. A. Morris, Senior Resident inspector -
                          J. D. Orr, Resident inspector
      Approved by:        James C. Linville, Chief, Projects Branch 3
                          Division of Reactor Projects
                                                    I
                                                                      l
        9002090222 980203
        gDR      ADOCK 05000354
                            PDR
 
  .
    .
                                                                    .-
                                                                                                      ;
    .-
                                              EXECUTIVE SUMMARY-'
                                          Hope Creek Generating Station
                                      - NRC Inspection Report 50 354/9710
          This integrated inspection report includes aspects of licensee operations, engineering,
          maintenance, and plant support. The report covers a sever week period of resident
        . Inspection.
          Operations '
        = Plant operators conducted an essentially error free startup, in spite of numerous emergent-.
          plant equipment deficiencies. Operators exhibited excellent control of planned evolutk'ns
          and responded ' adequately to unanticipated equipment malfunctions. Reactivity
          management during the evolution was good. (Section 01.1)
        : Operators generally exhibited good performance with respect to procedure compliance,
        ; peer checking, and communication during infrequently performed evolutions while
        : preparing for plant start up. . Weaknesses were evident with respect to technical .
          specification required equipment status control.' (Section 04.1)
                                                                                                        ,
H      ' Operating crew performance during a requalification exam simulator scenario was .
          adequate. Though crew communications and event recognition were good, consistent
          with observed performance in the actual conuoi room, operators failed to perform or
L        evaluate certain expected actions during event response. Fidelity between the Hope Creek
          simulator and the actual control room was adequate. Examination security measures were-
          good.-(Section 05.1)
          The inspectors continued to observe good overall performance with respect to quality-
          improvement measures and oversight in operations. The inspectors noted that the recent -        !
          n..snagement focus on operability determination backlog reduction was effective. -(Section
          07)
          Maintenance                                                                  .
          Maintenance department performance, particularly with respect to procedure development
          and usage, was weak._ Repeat examples of inadequate procedure development and
        : implementation were identified.' in at least two cases, associated with maintenance on the      4
          reactor core isolation cooling systnm and a control room ventilation system, the
        ' performance issues resulted in_ delays in returning these safety systems to an operable
          status.' (Section M1.1)-
          LA recently-developed surveillance procedure which implemented technical specification
          acceptance criteria for electrical protection assembly under-frequency testing was
          -inadequate. (Section E2.1)
                                                          ii
i
 
                        _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ --      ..          ..  .
.
.
  PSE&G personnel completed inservice testing of the "A" and "C" core spray pumps safely
  and in accordance with the established procedure. (Section M2.')          .
  Enoineerina
  Engineering support of safe plant operation was acceptable. Several modifications were
  developed and installed during the refueling outage which improved the overall material
  condition of the plant. However, the backlog of engineering work activities, particularly
  with respect to completion of corrective action program related activities, had not been
  significantly reduced. Specifically, the timeliness with which condition report evaluations
  were completed and corrective actions established exceeded program expectations.
  (Section E2.1)
  PSE&G efforts at reducing the number of installed temporary modifications was effective.
  Remaining modifications were properly evaluated, with proper revisions made in associated
  drawings and procedures. However, several of the remaining modifications had been in
  place well beyond that which was intended by PSE&G's administrative control program.
  (Section E2.2)
  PSE&G failed to establish adequate testing and procedure controls to ensure that the
  operability of the reactor building to-torus vacuum breaker assemblies was maintained.
  This issue was self-identified, properly reported, and promptly corrected. (Section E2.3)
  The material condition of the reactor core isolation cooling system and high pressure
  coolant injection system was poor as indicated by failed technical specification operability
  testing during the reactor plant start up. Inadequate engineering practices contributed to
  the failure of the reactor core isolation cooling system surveillance testing. (Section M2.1)
  Resolution of a potentially generic deficiency in safety-related swing check valves was
  timely and effective. (Section E8.1)
  PSE&G implemented timely corrective actions following self-identified and reported
  discrepancies in the safety auxiliaries cooling system. These discrepancies, which rangoo
  from insufficiently-Justified system operating procedures to errors in design basis
  assumptions, were identified during a comprehensive design and licensing basis validation.
  (Section E8.6)
  . Plant Suocort
  The emergency operations facility was well maintained and secured. (Section P2)
  PSE&G security personnel performed a thorough evaluation of an event involving an
  unauthorized protteted area access. Corrective actions were good. (Section S4)                  l
                                                                        ill
 
            .. --      __ _ _        _ -_
                                                                                                                                                                          ,
      . ,.
        .
                                                                      TABLE OF CONTENTS -
                    ,
                        EXECUTIVE SUMM ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
                    ' TABLE O F C C NTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv -
                      ' l . ' O pe r a t ion s . . . _ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -
                                              .
                                01        Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
                                            01.1 u Reactor Startup and Power Ascension . . . . . -, . . . . . . . . . . . . . . 1
                                04        Operator Knowlooge and Performance . . . . . . . . . . . . . . . . . . . . . . s . . . 2
                                            04.1 - General Operator Performance Observations . . . . . . . . . . . . . . . . 2
                                -05'        Operator Training and Qualification . . . . . . . . . .:. . . . . . . . . . . . . . . . 3
                                            05.1' Observation of Licensed Operator Requalification Exam . . . . . . . . 3
                                07_        Quality Assurance in' Operations . . . . . . . . . . . . . . . .-.. . . . . . . . . . . . . 4 ;
                                08        Miscellaneous Operations issue . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
                                            08.1 -(Closed) VIO 50-354/97-07-01: failure to promptly identify inoperable
                                                    electric fire water pump . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
                                            08.2 (Closed) LER 50 354/97 32: technical specification required shut
                                                  : down due to concurrent inoperability of the high pressure coolant
                                                    injection and reactor core isolation cooling systems . . . . . . . . . . . . 5
                        li . - M aint e n a nc e . . . . . . . . . . . . .- . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
                                                                                                                                                .
                                M1 -      Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . .' . . . . . . . . . 5
L                                          M1.1 General Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 -
                                M2'        Maintenance and Material Condition of Facilities and Equipment . . = . . . . . 7                                .
                                            M2.1 Maintenance and Surveillance of High Precsure Safety Systems . . .:7                                  -
                                            M2.2. Core Spray System inservice Testing Observation - . . . , . . . . . . . .--8
                                M8        . Miscellaneous Maintenance issuosi . . . . . . . . . . . . . . . . . . . . . . . . . . -. 9 -
                                          ' M8.1 (Closed) VIO 50-354/E96125-01013: Failure to plan appropriate ;
                                                    post-maintenance technical specification (TS) surveillance testing '. 9l
                                            M8.2 (Closed) VIO 50 354/E96-125-01023: Failure to conduct appropriate -
                                                    post maintenance technical specification surveillance testing . . . . 9.                          .
                                            M8.3 -(Closed) VIO 50-354/97-04-02:high pressure coolant injection (HPCI)
                                                      system feed water injection line inoperable . . . . . . . . . . . . . . . . 10 -                                  '
                                            M8.4 (Closed) LER 50 354/97 27:operaticn in a technical specification (TS)
                                                    prohibited condition due to improperly performed surveillances for -
                                                  : determining specific gravity of the 125 Vdc and 250 Vdc batteries
                                                      ..............................................10
                                            M8.5' l Closed) LER 50-354/97-28: technical specification (TS) prohibited
                                                  -
                                                      condition - f ailure to perform secondary containment isolation -
                                                      actuation system surveillances . . . . . . . . . . . . . . . . . . . . . . . . 10
                                            M8.6 (Closed) LER 50 354/97-29: operation in a technical specification (TS)
                                                    - prohibited condition due to removal of a residual heat removal (RHR) -
                                                      shutdown coeling suction line snubber . . . . . . . . . . . . . . . . . . . 11
                                          - M8.7 (Closed) LER 50-354/97-30: failed inservice test surveillance for
                                                      reactor building-to- torus vacuum breakers . . . . . . . . . . . . . . . . 11
    1
                                                                                        iv
  -        -                                                _
                  [
                                                                                                                                                                  ____J
 
    - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _                          _ _ _ _
                                                                                                                                .
                                                                                                                                                              .
                                                                                                                                                                            .                            .
                                                                                                                                                                                                                                .
  -
                ...
              1.                        .                                                                  -
                                                            .
'
                                                            s 111. Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1
                                                                                    E2-      Engineering Support of Facilities and Equipment ................. 11
                                                                                              E2.1    General Observations -. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
,
                                                                                              E2.2 Review of Temporary Modifications . . . , , . . . . . . . . . . . . . . . . .~ 13 -
L.                                                                                            E2.3 Inadequate Testing of Reactor Building toLTorus Vacuum Breakers-
                                                                                                        ..............................................15
                                                                                    E8      Miscellaneous Engineering issues ". . . . . . . . . . . . . . . . . . . . . . . . . . . 17
                                                                                                                                                    .
                                                                                              E8.1 .(Closed) P2150 354/9511: deficiency in Anchor Darling swing -
                                        <<                                                            ' chec k valves . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . 17
                                                                                          - E8.2 - (Closed) VIO 50-354/96-07-01:two examples of a failure 'o evaluate-
                                                                                          _.          changes made to the facility in accordance with 10 CFR 50.59 . .17
                                                                                          ^ E8.3- (Closed) VIO 50-354/E96125-02013: failure to promptly identify and .
                                                                                                      .. correct a condition adverse to quality . . . . . . . . . . . . . . . . . . . . 18
                                                                                          T E8.4 (Closed).VIO 50 354/E96-125-02023:f ailure to promptly identify and
                                                                                                      correct a conditlan adverse to quality . . . . . . . . . . . _ .-. . . . . . . 18
                                                                                              E8.5- -(Closed) VIO 50 354/E96-125-03013: failure to obtain NRC approval
                                                                                                      before making a change to the facility 1. . . . . .. . . . . . . . . . . . . . . 18
                                                                                              E8.6 (Closed) LER 50 354/96 22: condition alone that could have
                                                                                                      iprevented removal of residual heat - safety auxiliaries cooling system
                                                                                                        (SACS) deficiencies . . .= . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
                                                              1  1V. Plant Support - . . . . . . . . . . . . . - , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - . . . 2 0 :
                                                                                    P2_-    Status of Emergency Preparedness Facilities, Equipment, and Resources'
                                                                                              ...........-........>...............................20
                                                                                    S4      Security and Safeguards Staff Knowledge and Performance . . . . . . . . . . 20
                                                            - V. Management Meetings '. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
                                                                                    X1      Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . : 21.
                                                                                                          .                                                                                                                                _
                                                                                                                                                                                                                                            \
                                                                          -
                        _                                          _                  _        __                                      __
                                                                                                    _
                                                    -- i
  ,.
                                                                                                                                    v
h'
              _
                                                                                                                                                                                                    -
                                                                                                                                                                . . . , . . . . . _ _ . . . . . - . . . _ , _ _ - - - - - - _ _ - _ _ _ _ _
 
  _ _ - - _ .
              . _ - _ _ _ - _ _ _ _ _ - _ _ _ - _ _ _ _ - _ _ _ _ _ - -
  .
  .
                                                                              Egnort Details
                                                                              1. Operations
                01                        Conduct of Operations
                Qld Reactor Startuo and Power Ascension
                    a.                  Insnection Scoce (71707)
                                          The inspectors observed Hope Creek restart activities from November 30,1997 to
                                          December 8,1997. Specifically, the inspectors monitored the reactor start up to
                                          criticality, extended reactor plant operation in Operational Condition 2, low pressure
                                          testing of the high pressure coolant injection (HPCI) system, a power and pressure
                                          reduction from 900 psig to 100 psig to resolve discrepancies with a jet pumo
,                                          operability surveillance, and safety relief valve lift testing.
!
                      b.                  Observations and Findinos
f
'
                                          In general, operator activities conducted during the plant start up were safe and
                                          deliberate, procedure adherence was very good, command and control was evident,
                                          and the control room atmosphere was quiet and unobstructed. The operators
                                          anticipated upcoming evolutions and questioned unexpected results. The inspectors
                                            noted that there was continuoc eactor engineering personnel coverage during
                                          reactivity manipulations. Apprwate decisions were made after equipment
                                          deficiencies were identified. As an example, the control room operators,
                                            anticipating a change to Operational Condition 1 from Operational Condition 2,
                                            performed the " Recirculation Jet Pump Operability - Daily (HC.OP-ST.BB OOO1(Q))"
                                            surveillance to verify that the average power range monitor (APRM) recirculation
                                            flow units were reading accurately. Rasults of this surveillance did not meet the
                                            acceptance criteria. Had the operators not performed this surveillance early, they
                                            would have been challenged with a potential inoperability of the APRMs in
                                            Operational Condition 1. This anomalous condition was promptly resolved before
                                            startup activities proceeded.
                                            However, the operators were challenged with several equipment issues that delayed
                                            the plant start up and required abnormal plant start up operations. For instance, the
                                            initial reactor start up wa. delayed to resolve a higher than normal reactor water
                                            cleanup (RWCU) system differential flow. The operators questioned the abnormal
                                            indication when the RWCU system was restored to service after outage
                                            maintenance to repair a piping flange leak. A portion of the RWCU was taken out
                                            of service and drained for a boroscope inspection of a flow measuring orifice.
                                            PSE&G engineers suspected, based on system parameters, that the flow orifice may
                                            have been installed backwards during the refuel outage maintenance activities. The
                                            boroscope inspection confirmed that the orifice was reversed. (See Section M1.1
                                            for details)
                                            On December 5,1997, five days into the plant start up, while performing reactor
                                            core isolation cooling (RCIC) system testing at rated plant pressure, control room
                                            operators discovered that the turbine speed controller had no effect on steam flow.
                                                                                                                                  ___n
 
                        _ ___ ___-_-____ - ___    - _ - - -_
  .
  .
                                                      2
          The RCIC system was declared inoperable. Control room operators performed the
          high pressure coolant injection (HPCI) system testing at rated plant pressure later
          that same day. After 24 minutes of operation, the operators tripped the HPCI
          turbine due to high vibrations. With both the RCIC and HPCI systems inoperable,
          operatore entered technical speeltication 3.0.3 and commenced a plant shutdown to
          iess than 150 psig as required. PSE&G personnellater determined that both the
          RCIC and HPCI malfunctions were related to earlier outage sctivities on those
i        systems. (See Section M2.1 for details)
f
)
l    c.  Conclusions
          Plant operators conducted an error free startup, in spite of numerous emergent plant
          equipment deficiencies. Operators exhibited excellent control of planned evolutions
          and they responded adequately to unanticipated equipment malfunctions. Reactivity
          management during the evolution was good.
    04    Operator Knowledge and Performance
    .QL1 General Ooerator Performance Observations
      a.  inspection Scone (71707)
          Throughout the report period, the inspectors observed numerous operations
          department activities both in the control room and in the field. Several tours with
          equipment operators were conducted to evaluate non-licensed operator
          performance,
      b.  Observations and Findinas
          Station operators performed several infrequent evolutions during the report period
          while preparing the plant for restart from the refueling outage. The inspectors
          observed generally good performance with respect to procedure compliance, peer
          checking, coordination and communication with other departments, and use of
          shutdown risk assessment information. As an example, when it became evident
          that the impending plant start up would be performed with only two of the three
          reactor feed pumps available, operators evaluated the viability of plant operation
          with only two feed pumps during a " test" start up in the Hope Creek simulator.
            Good command and control of a post-refueling reactor vesselinservice leak rate test
          was evident. Control rod scram time testing and control rod drive mechanism
            " friction" testing were completed in accordance with applicable procedures and
            without incident.
            The inspectors questioned operators regarding specific instances of inadequate
            safety-related equipment status control. For example, on December 2,1997, the
            inspectors determined that the technical specification (TS) action statement tracking
            log, maintained by the shift supervisor as a ready reference of the status of all TS-
            related equipment, did not accurately account for the "B" primary containment
            hydrogen / oxygen analyzer being out of service for maintenance. Other similar
                                                                                                  i
                                                                                                  l
 
      _ _ _ __ _. _ _ _._.__._. _ _._ _ . _ _ _                                              _        -
                                                                                                          . _ _ _ _ _
.        .
        .
                                                                                                                          l
                                                                    3                                                    :
                                                                                                                        '
,                        examples of.this weaknese had been noted previously by NRC inspectors with
                      . respect to the operability of the standby liquid control system. Despite the
}                        weaknesses in maintaining the status log, the inspectors determined from narrative
                        log reviews, operating shift turnover sheets, and individual Interviews that operators          ,
                        were aware of the actualinoperable system conditions. This minor violation of TS                !
                        6.8.1.a, which requires administrative procedures be implemented for equipment                  i
                        status control, is being treated as a Non Cited Viol? tion consistent with Section IV          j
                        of the NRC Enforcement Policy. (NCV 50 354/97<10 01)
      -
                                                                                                                        l
                                                                                                                        i
                        On November 19,1997, the inspectors noted that the reactor pknt mode switch in                  !
;                        the control room was in " Refuel," permitting single control rod witndrawals, in spite          l
?-                      of the fact that no rod withdrawals had been conducted for nearly twelve hours.                l
                        Operators had temporarily interrupted scheduled contro! rod testing to address other            '
                        emergent concerns, but elected to leave the modo switch in " Refuel" in part
                      - because the duration of the rod testing delay was indeterminate, and also because
                        returning the modo sw tch to " Shutdown" would gsnerate an undesired scram
i,                      signal. While no consequence resulted from this issue, the inspectors judged that            -l
                        the operators' decision to remain in the " Refuel" mode for an extended period was    _      1
                                                                                                                      '
                        not consistent with PSE&G management expectations for conservative operating                    .
                        practices.
                                                                                                                        l
                                                                                                                        .
              c..        Conclusions                                                                                    [
                        C: erstors exhibited good performance with respect to procedure compliance, peer                3
                        checking, and communication during infrequently performed evolutions while                      :
                        prepaling for plant start up.' Weaknenes were evident with respect to techical-
                        specification required equipment status control.                                                .
          05'          Operator Training and Qualification
          QLj, Observation of Licensed Onerator Reaualification Exam
            s..          Inanaction Scone (71707)                                                                      l
                        The i,.;pectors observed ons operating crew during a routine licensed operator
                          requalification examination in the Hope Creek control room simulator. Crew:
                          communications, procedure usa;,e, and control board operation were evaluated.
                          Additionally, simulator f_idelity to the actual control room configuration was ~
                                                                                                                        '
                          assessed. Lastly, the inspectors witnessed PSE&G instructor performance during -              ;
                          the conduct of the exam scenario,                                                              ,
                                                    .-                                                                  !
    v
  ..
 
                                    _ _ _ _ _ _ _ _ _ _ . ..
            .
                                                                  .
                                                                    ..    .  .    . ._.                .. .
    *
      ,
    4
                                                              4
,
        b.  ' Observations and Findinas .
                  On December 19,1997, the inspectors observed the conduct of a routine licensed
                  operator requalification exam simulator scenario. Security measures to ensure that
              . examination materials were not comprised prior to or during the examination were
                  . good. The inspectors did not detect any incidences of PSE&G evaluators "cuelng"
                                                                .
                  the examinees. Simulator fidelity with the actual Hope Creek control room was
                  adequate.- Recently implemented control room modifications were noted to be
  .
              . duplicated in the simulator, however emergency operating proccdures were " pre-
                  staged" at the shift supervisor desk prior to exam commencement, and well before
;                  conditions developed which would have required their use.
                    During the exam scenario, the inspectors observed that operators failed to insert a
l                  " half scram" signal in the reactor protection system following an event which
'
                    should have automatically cour,ed .s half scram signal but did not. The inspectors
                    learned that the evaluators also detected this failure, which led to their conclusion -
                    that the operating crew was * weak" overall in this scenario. Additionally, the
                    inspectors identified inconsistencies in plant operating procedures with respect to      ;
                    recirculation system operation. Specifically, the operating crew intentionally            '
                    reduced reactor recirculation flow to a minimum prior to inserting a manual scram,
                    tyhich resulted in a brief period of plant operation in the power / flow *instsility
                    region." While the practice of minimizing recirculation flow prior to a planned scram
                    was in accordance with the procedure employed during the exam scenario, other
                    operational guidance specifies that recirculation flow never be intentionally reduced
              . into the instability region.- Crew communications were good during the exam
                  l scenario, consistent with observed performance in the actual control room during
                    plant events.-
              - The inspectors discussed their findings with the Hope Creek' operations manager
                    following completion of the exam. This individual expressed agreement with the
              inoted observations, and planned to address them appropriately.
        _c.        Conclusions -
                '
                : Operating crew performance during a requalification exam simulator scenario was
                  -adequate. Though communications and evtnt recognition were good, consistent
                    with observed performance in tM actual control room, operators f ailed to perform or
                    evaluate certain expected actions during event response. Fidelity between the Hope
              . Creek simulator and the actual control room was adequate. Examination security
              . measures were good.
        07          Quality Assurance in Operations
                [Ehe inspectors continued to observe gbbd overall hrformance with respect to
                  -quality improvement measures and oversight in operations. The inspectors
                    reviewed quality assurance department reports, interviewed auditors, and attended
                    an offsite Nuclear Review Board meeting. These activities provided effective
                    performance assessment feedback to operations management, and in a timely and
                    useful manner.-- Performance trending in various categories was also evident.
              _
 
                  _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .                _.
  .
  .
                                                        5
          The Inspectors noted that a recent management focus on operability determination
          backwg reduction was effective. At the conclusion of the refuel outage, only eight
          operability determinations remained from the initial pre outage population of thirty-
          five. = Appropriate safety evaluations were completed for the items which remained
          to ensure that continued operation with degraded componentn did not pose a threat
          to plant safety.
    08    Miscellaneous Operations issue
    .QHJ IClosed) VL a354/97 07-01:f ailure to promptly identify inoperable electric fire
            water pump. ihe inspectors reviewed PSE&G's letter dated December 12,1997
            which provided a response to this violation. PSE&G attributed the cause of this
            event to personnel error, in that operators failed to follow a procedure precaution
            during an electrical bus swap. Further errors were committed by both operations
            and fire protection personnel in that these individuals failed to detect the condition
            during localinspection tours of the equipment. The inspectors verified that PS2&G
            completed the corrective actions stated in the noted letter, which ;ncluded individual
            disciplinary action, additional training, and procedure enhancements. The fira water
            pump was promptly returned to an operable status upon discovery of this issue.
    W (Closed) LER 50 354/97-32: technical specification required shut down due to
          - concurrent innoerability of the high pressure coolant injection and reactor core
            isolation cooling systems. This issue is described in detail in Section M2.1 of this
            report. No new information was provided in this LER.
                                                  11. Maintenance
1
!  ; M1.  Conduct of Molntenance
    ML1 Deneral Observations
      a.  Inspection Scope (62707)
            The inspectors observed or reviewed numerous maintenance department activities
            during the report period, which included outage work practices and plant startup
            testing evolutions.
      b.  Observation and Findinas
            At the conclusion of the refueling outage and during the reactor start up, the
            inspectors noted several maintenance department performanco deficiencies which
          -impacted plant operation. For example, the inspectors conducted a thorough
            cor.tainment close-out inspection tour prior to plant start up, and identified several
            items left in the drywell by maintenance personnel, including tools, loose insulation,
            and welding rods. This indicated a lack of sensitivity on the part of maintenance
            technicians and supervisors regarding the need to minimize foreign materialin this
            area, particularly in light of the fact that the emergency core cooling suction strainer
            modification was not fully implemented during the outage.
                                                                                                      (
                                                                                                      l
 
      _ _ _ _          ___ __.
    .
    .
                                                      6
          On November 24,1997, operators experienced two automatic trips of the BK400
            safety related control room venthtlon system chiller unit due to high condenser
          backpressure. Subsequent PSE&G investigation determined that a recent
          maintenance activity to calibrate a system pneumatic pressure controller was not
          properly completed. Specifically, a gain adjustment in the control circuit for the
          condenser cooling water pressure control valve was laft in the wrong position              '
          following the calibration, resulting in insufficient coolir.g to the unit during operation.
                                                                                                      ~
            Contrary to PSE&G requirements, "as found" data was not recordd, and procedure
              steps were not initialed as being satisfactorily completed. Additionally, the
            technician involved in the worL activity was neither trained not qualified to work on
            pneumatic controllers. Lastly, supervisory review of the completed work package
              failed to detect any errors. This lack of attention to detail and failure to follow
              established procedures while performing maintenance on a safety related
              component was a violation of technical specification '3.8.1.a. (VIO 5C 354/9710-
              02)
              On November 29,1997, based on anomalous control room indications, PSE&G
l
              engineers determined that a orifice plate flow element in the reactor water cleanup
              (RWCU) system had been installed backwards following system piping flange leak
              repairs. Rather than reverse the orifice to its proper orientation, engineers
              completed a design modification and safety evaluation which re calibrated the
              associated flow indication circt.it to account for the new configuration. During
f            subsequent reviews into the causal f actors of this issue, the inspectors determined
              that the work order guidance provided to the technicians performing the RWCU
              piping flange repairs failed to specify the correct orifice plate installation.
              Additionally, the technicians did not demonstrate quality workmanship during the
              activity in that the "6s found" orifice orientation was not recorded as an aid for
              reassembly. This failure to establish appropriate instruction for maintenance on
              components which can affect to performance of safety-related equipment,in this
              case RWCU isolation controls, was a violation of 10 CFR 50 Appendix B, Criterion
              V. (VIO 50 354/9710 03).
              On December 30,1997, while attempting to restore from on line reactor core
              isolation cooling (RCIC) system maintenance, technicians were unable to open the
              turbine trip throttle valve. Subsequent investigation, which include discussions with
              appropriate vendor representatives, determined that the mechanical overspeed trip
              device was improperly reset which in turn resulted in the inability to reset the
              turbine trip throttle valve. Neither the RCIC system operating or maintenance
              procedure adequately described the proper method to ensure this interaction was
              minimized or precluded oilor to returning the system to standby service. The
              inspectors noteti that this problem was not previously understood in spite of
              previous industry experience, including a similar 1993 issue with auxillary feedwater
              pump turbines at the Salem station. The f ailure to establish appropriate procedures
              for the RCIC system as a violation of technical specification 6.8.1.a. (VIO 50-
              354/07 10-04)
              Lastly, the inspectors identified an improperly developed TS surveillance test
3              procedure for electrical protection assemblies. This issue is described in detai! in
              section E2.1 of this report.
  i
 
.
f
                                                7
  c.  Conclusions
        Maintenance department performance during the report period, particularly with
        respect to procedure development and usage, was weak. Three examples of
        inadequate work controls were identified, which resulted in delays in returning
        safety systems to an operable status.
  M2-  Maintenance and Material Condition of Facilities and Equiprnent
  M2d Maintenance and Surveillance of Hioh Pressure Safety Systems
    a.  inspection Scooe (61726,62707)
        During the plant start up following RFO7, the inspectors observed surveillance
        testing and maintenance associated with the high pressure coolant injection (HPCI)    -
        system, the reactor core isolation cooling (RCIC) system, and the automatic
        depressurization system safety relief valves (SRV).
    b.  Observations and Findinog
        Station operators experienced several equipment failures with high pressure safety
        systems during technical specification-required surveillance testing. For example, in
        spite of a high quality pre evolution briefing prior to 800 psig SRV " lift" tests in
        Operational Condition 2, difficulties were encountered with the set up of measuring
        and test equipment which required operators to re perform the test. Additionally, a
        faulty acoustic monitor was identified on the "C" SRV which required replacement
        and ratesting. The tests were conducted with good management oversight of the
        activitica and effective monitoring of critical plant parameters.
        On December 5,1997, the RCIC system was declared inoperable when control
        room operators discovered that turbine controls were ineffective at changing turbine
        speed. PSE&G later determined that the governor valve stem was binding in the
        valve body packing gland. A detailed root cause investigation into this event, wnich
        was effect vely coordinated between operations, maintenance and engineering
        departmems, determined that the valve stem b!nding was primarily the result of
        inadequate engineering review of an earlier March 1997 design change which
        replaced the RCIC governor valve stem with a stem of a different material. This
        modification was implemented as a corrective action to earlier governor valve stem
        f ailures in December 1996, when the RCIC turbine suffered an overspeed trip
        because of corrosion on the governor valve stem. PSE&G engineers, in concert
        with contracted turbine vendor personnel, determined that the new valve stem
        material exhibited c thermal expansic,n rate three times that of the original stem,
        which resulted in stem binding when subjected to rated reactor temperature and -
        pressure conditions. Additionally, manufacturing tolerances of the replacement
        stem were not equivalent to that of the original, exacerbating the issue. Corrective
        actions to restore the system focused on widening the tolerances of the governor
        valve packing gland spacers to accommodate in the increased thermal growth rate.
        The system was properly tested and restored to operable status on December 12,
.        _ - .
 
          - _ _ _ _ - _ _ _ - _      .
                                                              .
                                                                                                -
  .
                                                  8
          1997. However, inadequate corrective actions were implemented folic, wing the
        previous RCIC governor valve stem failure, in violation of 10 CFR 50 Appendix B,
        Criterion XVI " Corrective Action." (VIO 50 354/97 10 05)
        in addition to the December 5,1997 RCIC testing described above, operators
          condected HPCI system surveillance testing later on that same day. This test also
          f ailed due to high vibrations detected approximately 20 minutes into the test run.
          Since both HPCI and RCIC were declared inoperable while in Operational Condition
          2, operators entered technical specification (TS) 3.0.3 and reduced reactor plant
          power and pressure to the point at which neither sistem was required to be
          operable (l.a.< 150 psig). Operators stabilzod the plant in the appropriate condition
          safely and well within TS prescribed criteria. Again, a comprehensive team
          comprised of individualc from various departments and vendor representatives was
          assembled to develop a troubleshooting plan and determine the root cause of the
          problem. After a thorough review, which included intrusive system inspections, the
          team concluded that excessive moisture collection in the turbine glands corroded
          the internal carbon steam seals and caused the noted turbine vibrations. The
          inspectors reviewed the team's efforts and found that the team's conclusions were
          reasonable. Corrective actions included the replacement of the gland seal
          components, and the initiation of a system modification to provide improved gland
l
'
          seal cavity draining. The system was successfully tested and restored to operability
          on December 12,1997,
    c.  Conclusions
          The material condition of the reactor core isolation cooling system and the high
          pressure coolant injection sys'3m was poor as indicated by failed technical
          specification operability testing during the reactor plant start up, inadequate
          engineering practices contributed to the failure of reactor core isolation cooling
          system surveillance testing.
    M2.2 Core Sorav system inservice Testina observation
    a.    Insoection Scope (61726)
          The inspectors observed the performan:e of the "A" and "C" core spray pump in-
          service tests (IST). Other than the pre-evolution brief conducted in 'he control
          room, the inspectors focused on the actions of the operators and technicians in the
          local core spray pump rooms,
    b.    Observations and Findinas
          The pre-job brief for the quarterly inservice test was thorough and was conducted in
            accordance with the operations department standard checklist. A.ll participants in
            the evolution were present and actively contributed lessons learned and contingency
            plans for potential unexpected occurrences,
                                                                                                  i
                                                                                                  I
                                                                                                  l
                                                                                                  1
 
  ._.  _ _ _ _    _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - _ - _ _ - _  -
      .
                                                                            9
                                  The inspectors interviewed the cognizant equipment operator and instrument              '
                                  technician regarding the installation and use of their respective measuring and test
                                  equipment (M&TE). These individuals were knowledgeable in the use of the M&TE.
                                  The inspectors observed that the MGTE was properly installed and within
                                  calibration.
                                  The inspectors observed the instrument technician spill some potentially
                                  contaminated water from a system test connection. The technician reported the
                                  potential problem to the radiological controls technician when he arrived on scene to
                                    support the surveillance. The inspectors determined that their actions were prompt
                                    and reasonable to preclude any potential radiological problems,
                c.                Conclusions
                                    PSE&G technicians completed inservice testing of the "A" and "C" core spray
                                    pumps safely and in accordance with the established procedura and technical
                                    specification 4.0.5,
                M8                  Miscellaneous Maintenance issues
i
l              MB.1 (Closed) VIO 50 354/E96-125-01013: Failure to plan appropriate post maintenance
                                    techriical specification (TS) Surveillance testing. NRC inspectors identified that
                                    PSE&G had not planned to perform TS 4.1.3.2.b required control rod scram time
                                    testing before entering Operational Condition 1. This testing was required to
                                    demonstrate the operability of over sixty control rods which had undergone
                                    maintenance during RFO6. Similarly, control rod scram time testing had not been
                                    adequately performed prior to restart from earlier refueling outages, specifically
                                    RF03 and RF05 (see M8.2 below). The inspectors reviewed the corrective actions
                                    implemented as described in PSE&G's November 22,1996 violation response letter.
                                    PSE&G's actions immediately following identification of this issue were prompt and
                                    comprehensive. These actions included conduct of the proper TS surveillance
                                    testing, a review of other " conditional" TS requirements, and independent
                                    assessments of the station's readiness for a change in operational condition.
                                    Subsequently, in-depth TS training was provided to operations personnel to
                                    reinforce the nc ' to operate and maintain the plant in accordance with the
                                    licensing basis, Lastly, the inspectors reviewed PSE&G control rod testing plans
                                    following RFO7 maintenance. The inspectors verified that all required scram time
                                    tests were completed in the proper operational condition and in compliance with TS
                                    4.1.3.2.b.
                MM IChsed) VIO 50-354/E96-125 01023: Failure to conduct appropriate post-
                                    maintenance technical specification surveillance testing. This issue involved the
                                    discovery that following maintenance in RF03 and RF05 which had the potential to
                                      affect control rod scram insertion times, no TS 4.1.3.2.b surveillance testing was
                                      performed as required. The corrective actions for this issue are identical to those
                                      described for the event described in section M8.1 above,
                                                    w
 
      .  - -_            _ _ - .  _-        -    -.    -      --          . - - - .          .    -_ -
    .
    .
                                                        10
        MHJ (Closed) VIO 50 354/97 04 02:high pressure coolant injection (HPCI) system feed
                water injection line inoperable. This event, along with PSE&G's root cause analysis
                and proposed corrective actions to prevent recurrence, were described in detail in
                LER 50 354/9713 00and 01. The inspectors reviewed these documents and the
                stated corrective actions, and documented their findings in NRC Inspection Report
                50 354/97 09. Additionally, PSE&G responded directly to the violation in a
                September 9,1997 letter to the NRC. This letter did not provide any significant
                additionalinforrration beyond what was described in the LER and its supplement,
                with the exception that PSE&G disagreed with the inspectors' assessment that the          .
                " initial decision to maintain the operability of the HPCI system was non-
                conservative." While the inspectors acknowledged PSE&G's argument that
                operators made the decision in accordance with the " normal plant (operability
                                                                _
                determination) process," the inspectors maintained that the condition of the HPCI
              ' system feedwater injection valve was not certain at the time the decision was
                made, and that the organization did not effectively validate, verify, and interpret
                information gathered through troubleshooting.
        MHd (Closed) LER 50 354/97 27: operation in a technical specification (TS) prohibited
                condition due to improperly performed surveillances for determining specific gravity
                of the 125 Vdc and 250 Vdc batteries. TS 4.8.2.1.b.1 requires that specific gravity
                of the safety related batteries be determined every 92 days and be within the              -
                specifications provided in Table 4.8.2.1 1. TS Table 4.8.2.1 1 requires that each
                cell's specific gravity be corrected for temperature for every individual cell. Hope
                Creek previously used, since initial operation, every sixth cell to correct for
                temperature. The inspectors noted in the LER that this change in methodology
                produced very minor changes in specific gravity measurements. The inspectors
                verified by on site inspection that PSE&G completed procedure changes for the
                affacted surveillances. This licensee identified and corrected violation of TS is being
                treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC
                Enforcement Policy. (NCV 50 354/9710 06)
l      M{Lii (Closed) LER 501354/97 28: technical specification (TS) prohibited condition -
                failure to perform secor.dary containment isolation actuation system surveillances.
                This event involved :.he self identified discovery that outage planning personnel
                failed to include channel functional tests required by TS in the outage plan. As a
L              result, surveillance activities required by TS 4.3.2.1 were not performed to
                demonstrate operability of the secondary containment isolation achation logic
                channels during core alterations or during operations with a pot .al to drain the
                reactor vessel. PSE&G attributed this failure to personnel error in that outage
                planners inappropriately used an uncontrolled TS surveillance matrix while
                developing the refueling outage testing schedule. This matrix did not accurately
                reflect all TS surveillance testing requirements. PSE&G procedures require that the
L              contro!!ed surveillance testing matrix, which was recently validated by the technical
'
                specification surveillance improvement project, be used for planning and scheduling
                TS required testing activities. Corrective actions to preclude recucence of this issue
                involved Individual disciplinary measures and re emphasis on mapp;nent
                expectations regarding the use of only controlled documents. Based on an in-office
                review the inspectors concluded the corrective actions were adequate.
l
l
  .                              .    .                      .        . _ .          _.
 
          - _ _ - _ - _ _ _ _ _ _ _ -
  .
                                                    11
          This licensee identified and corrected violation of TS 4.3.2.1 is being treated as a
          Non Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy.
          (NCV 50-354/9710 07)
    MlLQ (Closed) LER EO 354/97 29: operation in a technical specification (TS) prohibited
          condition due to removal of a residual heat removal (RHR) shutdown cooling suction
            line snubber. On November 18,1997, during implementation of a design change
          package (DCP) to replace snubber 1 P BC 144 H002on the RHR shutdown cooling
            suction line, personnel identified that the snubber had been mistakenly removed in
            1992 during the snubber reduction program. PSE&G's investigation determined that
            individual work orders to remove snubbers were not generated from the approved
            snubber reduction DCP in 1992. PSE&G performed a comprehensive walkdown of
            the shutdown cooling suction line and did not identify any additional missing
            snubbers. PSE&G also performed a review of outage work orders and did not
            identify any activities that would have unintentionally removed snubbers.
            The inspectors interviewed licensee personnel in PSE&G's inservice inspection and
            work planning departments to determine the effectiveness of the corrective actions.
            All required system snubbers were replaced with a new model during RF07. The
            inspectors determined that appropriate licensee personnel v'ere knowledgeable of
            this event and that they had understood the lessons learned. This licensee
            identified and corrected violation of TS 3.7.5 is being treated as a Non-Cited
            Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-
            354/97 10 08)
    ML2 (Closodi LER 50-354/97-30: failed inservice test surveillance for reactor building to-
,
!
            torus vacuum breakers. This issue was discussed in detailin section M2.3 of this
            report. This LER accurately described the circumstances involved in the discovely
            and resolution of the issue.
                                              Ill. Enainee.ririg
    E2      Engineering Support of Facilities and Equipment
    E2d General Observations
    a.      Insocction Scoce (37551)
              The inspectors reviewed several engineering department work products, including a
              quarterly maintenance rule performance report, se,veral condition report root cause
              evaluations, and design change package safety evaluations. Intervicws were
              conducted with engineering personnel as appropriate.
    b.      Observations and Findinas
              During the report period, the PSE&G engineering department completed a
              reorganization which resulted in a staff reduction of nearly eighty employees.
                                                                                                  _______ _
 
.
                                            12
    Additionally, the department completed a comprehensive " engineering backlog"
    review which was a systematic assessment of all outstanding PSE&G engineering
    work items in ordar to prioritize necessary work activities and eliminate non-
    productive efforts. The inspectors concluded that this latter activity was a positive
    initiative to fully understand the volume of needed enginecting work following the
    reorganization. Further, based on personnelinterviews with operations and
    maintenance department staff, the inspectors found that these personnel were
    sometimes unaware of the appropriate engineering points-of contact.
    The inspectors noted that the backlog of corrective action program activities
    assigned to the engineering department continued to exceed programmatic
    expectations. Additionally, two condition report evaluations reviewed by the
    inspectors were not completed until well after programmatic time constraints.
    These two condition reports involved: (1) a question regarding the design adequacy
    of an unmonitored turbine building battery ventilation exhaust path, and (2) a f ailure
    of a solenold-operated valve in the MSIV sealing system.
    The inspectors reviewed the effectiveness of some of the design change packages
    (DCP) impleme'ited during the refueling outage. For example, the reactor
    recirculation pump seal modification, in conjunction with modifications which
    improved Leal drain lines, appeared to be effective in that initial seal cavity pressure
    trends during the first month of operation after the outage were much improved
    over the old design. Additionally, the unidentified leakage collected in the drywell
    floor drain sumps exhibited significant improvement,largely due to this modification.
    The drywell equipment drain sump identified leak rate also improved. Plant
    operators benefitted from other engineering department initiated modifications,
    including elimination of the rod sequence control system, removal of recirculation
    pump motor generator set ventilation runback signals, and installation of 4160 VAC
    switchgear panel test points. Virtually every mechanteal mubber in the facility was
    replaced with a newly designed and engineered hydraulic design.
    The inspectors performed a comprehensive review of DCP 4HE-0382 which
    replaced logic cards in the reactor protection system (RPS) end power range neutron
    monitoring system (PRNMS) electrical protection assemblies (EPA). This
    modification was developed based on vendor recommendations. The 10 CFR 50.59
    applicability review associated with the DCP was sufficiently detailed and properly
    referenced the necessary licensing basis documents to permit the inspectors to
    independently reach the same conclusion es PSE&G, specifically that the
    modification did not require a safety evaluation.
    In conjunction with this DCP evaluation, the inspectors reviewed the completed
    work orders and installation and test procedures utilized to complete the logic card
    modification. The work order documentation was sufficiently detailed, and all
    necessary approvals, tag outs, and retests were specified. The post maintenance
      test required successful completion of procedure HC.MD ST.SB 0000(Q) Revision 0,
      ~16 Month Electrical Protection Assembly Channel Calibration." Two deficiencies
      vvere identified as a result of the inspector's review of this completed procedure.
      Specifically, maintenance technicians failed to document "as found" voltage and
                                                                                                i
  _                                                                            _.          ..
 
          __-____ - _ _
.
.
                                                  13
        frequency readings to et least the same number of signibcint digits 3 stated in the
        procedure test acceptance criteria, meaning that the recorded data could not be
        adequately evaluated for acceptability. Secondly, step 5.9.8 of the procedure
        specified the TS acceptance criteria for under frequency measurements as "2 57
        hertz." The actual TS 4.8.4.6.b criterie for the PRNMS EPA's is "57 hertz 0%
          + 2%." Though "as lef t" setpoints were within the actual TS criteria, the f ailure to
        establish surveillance test procedures consistent with TS requirements waa a
        violation of TS 6.8.1.d. (VIO 50 354/9710 09)
        Additionally the inspector noted that the reactor core isolation cooling (RCIC)
        system surveillance test failure on December 5,1997, was primarily attributed to
        inadequate development of corrective actions to previous RCIC governor valve stem
        issues. (See Section M2.1 of this report for details)
  c.    Conclusions
        Engineering support of safe plant operation was acceptable. Severa! modifications
        were developed and mstalled during the refueling outage which improved the overall
        material condition of the plant. However, the backlog of engineering work
        activities, particularly with respect to completion of correctivt action program
        related activities, was not significantly reduced. Engineering support with respect
        to developing efffective corrective actions to previous reactor core isolatinn cooling
          system f ailures was inadequate. A surveillance procedura which implemented
          technical specification acceptance criteria for electrical protection assembly under-
          frequency testing was inadequate.
  E2,2 Review of Temocrarv Modifications
      a.  IDipaction Scone (37551)
          The inspectors reviewed the status and tracking of temporary modifications to
          verify that they met regulatory and PSE&G requirements for acceptability.
          Additionally, the inspectors conducted a thorough evaluation of one temporary
          modification, which included interviews with the applicable system manager and
          reviews of affected drawings and procedures.
                                                                                                r
      b.  Observatione and Findinos
          Hope (' reek had 24 active temporary modifications prior to the beginning of RFO7.
          By the end of the outage, only eight modifications remained, indicating a good
          management focus on temporary modificction reduction over the period. However,
          of the eight, five were greater than one year old. While appropriate safety
          evaluations had been completed Justifying the continued existence of each
          remaining modification, the inspectors determined that the intent of PSE&G's
          nuclear administrative procedure NO.NA-AP.ZZ-0013 (NAP-13), "Controlof
            Temporary Modifications," hbd not been satisfied. This procedure 3tipulates that
          such modifications should be "of short duration," and " generally not used for longer
          than 90 days." The oldest temporary modification at the facility was greater than
                                                                                      .
 
  _-    _ _ _ _ _ _ _ _
                                  ___-__- - -
.
                                                          14
                  three years old. All eight modifications had planned removal dates before or during
                  the next refueling outage.
                  Additionally, the inspectors noted that on December 12,1997, a quality assurance
                  inspector identified numerous minor deficiencies in the temporary modification
                  tracking log in the Hope Creek work contro! center. While none of the
                  discrepancies affected the operability of r.ny safety-re' .ted systems, they
                  collectively indicated tnat PSE&G's programmatic requirements were not being
                  satisfactorily implemented. The condition report documenting this issue thoroughly
                  described the deficiencies.
                  The inspectors performed a detailed assessment of tempor . / modification #94-
                  026, titled " Drain KP (MSIV Sealing) System Water from KL (PCIG) system," which
                    was installed on September 10,1994. This modification was installed tc mitigate
                    the effects of leakage from the main steam system into the normally idle MSIV
                    sealing system. This leakage was negatively affecting the operational performance
                    of the primary containment instrument gas (PCIG) system, which supplies
                    pneumatic control for several safety-related applications. The modification involved
                    installation of a drain hose to an MSIV systern test connection, and the opening of
                    normally deenergized solenoid operated va: /es.                                                      /
                    The 10 CFR 50.59 safety evaluation associated with this modification was
                    adequate; however, in part because of when it was written, it did not meet current
                    PSE&G engineering department expectations for quality. As a result of inspector
                    questioning, the evaluation was re written (and re approved by the station
                    operations review committee) using the current engineering department standards.
                    MSIV Sealing and PClG system drawings we.e properly revised to reflect installation
                    of the modification. Affected operating procedures were also revised appropriately.
                    The inspecto s asked why this modification had been permitted to endure beyond
                      NAP 13 expectations, and learned that initial attempts to resolve the issue with                    ;
                    corrective maintenance were unsuccessful. Later, a decision was made to remove                      '
                      the MSIV Sealing system entirely, but licensing actions were first required to effect
                    the change. Ultimately, this licensing action was deferred to the current operating
                      cycle due to PSE&G management placing greater priority on other needed licensing
                      actions.
    c.              Conclusions
                      PSE&G efforts at reducing the number of installed temporary modifications was
                      effective. Remaining modifications were properly evaluated, with proper revisions
                      made in associated drawings and procedures. However, several of the remaining
                      modifications had been in place well beyond that which was intended by PSE&G's
                      administrative control program.
                                                                                                                            l
                                                                                                            _ _ _ _ - _ _
 
    _ _ _ _ _ _ _      _______ - _-_ _ _                    _-
  .
                                                                                15
                  E2J inadeauste Testina of Reactor Buildina to Torus Vacuum Breakers
                  a.              Insocction Scoce (61726,37501)
                                  The inspectors reviewed PSE&G's actions to address insufficient reactor building to-
                                                                                                                              '
                                    torus vacuum breaker (RBTVB) testing which directly impacted their operability.
                                    Specifically, excessive back-leakage tnraugh instrument air (IA) system supply
                                    check valves would have prevented an accumulator, designed to provide a minimum
                                    of two hours of operation for the pneumatically operated RDTVBs, from performing
                                    its intended safety function. A two hour period is assumed in design Sosis
                                    calculations to provide adequate time for manual alignment of the backup pneumatic
                                    supMy from the primary containment instrument gas (PCIG) system,
                  b.                Observation and Findinas
,
                                    The RBTVBs are designed to open during certain accident conditions to prevent a
                                      subatmospheric condition from developing in the torus air space. Under certain
                                      circumstances, a small differential pressure would develop between the reactor
                                      building and torus, which could challenge the structuralintegrity of Me torus. Eithei
l
'
                                      of two sets of RBTVB would open to equalize pressure, in the Hope Creek design,
                                      there are two sets of RBTVBs (each comprised of s check valve and an air operated
                                      isolation valve (AOV) in series). An accumulator associated with each RBTVB
                                      assembly is designed to provide the capability of opening the RBTVB AOVs
                                      (1GSHV 5029 & 5031) over a two hour period when offsite power is lost. Each
                                      accumulator is supplied by two diUerent systems, the lA system (normal) and the
                                      PCIG system (backup), via isolation check valves. The PCIG supply piping also
                                      includes a normal ly closed manualisolation valve. The lA check valves (1GSV-081
                                        & 1GSV 093) must provide a relatively leak tight boundary to ensure that the
                                        RBTVB accumulator can perform its intended safety function.
                                        On Novt,mber 12,1997, P3E&G engineers discovered that previous testing of these
                                        lA supply check valves was inadequate since it verified only that the valves were
                                        closed by monitoring for gross leakage from the system, in order to preserve
                                        RBVTB design basis assumptions, proper testing should have periodicall/ quantified
                                        actualleak rate conditions. This would ensure that any lA check valve leakage
                                        would not reduce the RBTVB accumulator pressures to unacceptably low levels
                                        within two hours following a loss of offsite power. At the time of this discovery,
                                        the reactor plant was shutdawn for a refueling outage.
                                        The allowable check valve leak rate was derived from PSE&G calculation H 1 GS-
                                        MDC-1105, which demonstrated that the accumulator size of 19.0 cubic feet was
                                        sufficient to ensure one stroke of the AOV since it needed a minimum of 17.01
                                        cubic feet of _ air at 80 psig to operate. Thus, the maximum allowable leak rate
                                        through the interfacing check valves for the two hour design period was calculated
                                        to be 3345 secm. On November 21,1997, a leakage rate test was performed to            i
                                        measure the actualleakage through the lA check valves in both RBTVB trains. The
                                        inspectors noted that both trains exhibited leakage above the calculated maximum
                                        rates which challenged the operability of the RBTVBs. PSE&G initiated prompt
                                                                                                                                I
                                                                                                                            _ _j
 
  _.  .                                                                    _                  _.
.
.
                                              16
      actions to rcwork the seating surf aces of the deficient check valves, which was
      proven successful by subsequent satisf actory leak rate tests. In addition to the
      check valve rework and testing, Licenseo Event Report (LER) 07 030 00,lmproper
      in Service Test Hesults in Inoperable Containment Atmospheric Control System
      Vacuum Breaker Assemblies, indicates that the PSE&G testing program has been
      upgraded to list the affected valves as requiring leakage rate testing overy two
      years. The inspector verified that these actions were completed.
      Additionally, the Hope Creek design basis assumes that a backup pneumatic supply
      from PCIG will be manually aligned to the RBTVBs within two hours of the initial
      loss of offsite power. However, this requirement was never incorporated inn
      operating procedures. The inspectors judged that, as a result of the "as found" lA
      supply check valve leakage, combined with the lack of operational guidance to align
      a backup pneumatic supply, it was unlikely that the RBTVBs would have operated
      beyond one hour following the design basis events for which they were instr.: led to
      mitigate. This was a violation of 10 CFR 50, Appendix B, Criterion V which
      requires the establishment of procedures for activities that affect quality. These
      procedures must include appropriate quantitative and qualitative acceptance criteria
      for determining that important activities have been satisfactorily accomplished.
      These circumstances, which resulted in the simultaneous degradation of two
      independent and redundant trains of safety related primary containment vacuurn
      relief, both required to be operable per technical specification 3.6.4, have existed
      since initial plant construction and operation. Appropriate testing, if conducted,
      would have confirmed the existence of excessive lA check valve leakage which
      would have necessitated repairs to the affected valves. This would have provided
      reasonable assurance that the RBTVBs would have been maintained in an operable
      status fer the two hour period specified in the design basis. Thus, PSE&G's self-
      identified failure to perform required leakage rate tests of the lA supply check valves
      associated with the RBTVB assemblies was a violation of 10 CFR 50, Appendix B,
      Criterion XI, Test Control, which requires appropriata testing of systems bo
      identified and conducted in accordance with applicable procedures that specify
      acceptance limits derived from design documents. LER 97-030 notes that operating
      procedures have been revised to include operator actions to place the backup
      pneumatic r,upply in service to support RBTVB operability. The inspectors verified
      that these actions were comp eted.
      These licensee identified and corrected violations of 10 CFR 50 Appendix B, Criteria
      V and XI, are being treated as Non Cited Violations, consistent with Section Vll.B.3
      of the NRC Enforcement Policy. (NCV 50 354/97-1010and 971011)
  a.  Conclusions
      PSE&G failed to establish adequate testing and procedure controls to ensure that
      the operability of the reactor building to-torus vacuum breaker assemblies was
      maintained. Specifically, calculational assumptions made to encure their operation
      following design basis events were not adequately translated into plant operating
      and testing procedures. However, this deficiency was self identified during a non-
 
        - _ _ - _ ___ __ _____ _ _ - - - _                                                                                                    .
  .
  .
                                                                  17
                        routine comprehensive inservice test program review, and resulted in the prompt
                        development and imp!ementat;on of appropriate corrective actions.
    E8                  Miscallaneous Engineering issues
    fil [ Closed) P2150 354/9511: deficiency in Anchor Dariing swing check valves. This
                        potentially-generic issue, and the events which led to its discovery at the Hope
                        Creek station, were discussed in detailin NRC Inspection Reports 50 354/94 26
                          and 95 03, PSE&G made an initial NRC notification of this issue, which involved a
                          design and configuration control deficiency with sixteen pump discharge check
                        valves (twelve safety related), on March 29,1995. PSE&G subsequer:tly completed
                          a formal 10 CFR 21 notification in a letter dated April 28,1995. Because of the
                          reported deficiencies, the moving valve disks and hinge arms were contacting and
                          deforming the valve bodie . The inspectors reviewed the corrective actions
                          proposed in PSE&G's report document, v'hich included internalinspections of each
                          susceptible check valve. Based on the inspections, several hinge arms and disks
                          were replaced or repaired by weld buildup. The inspectors verified by on site
                          inspection that internal valvo inspections completed subsequent to these repairs
                          proved that the repairs were effective in preventing further valve damage.
1
    EfL2 (Closed) VIO 50 354/96 07 01:two examples of a failure to evaluate changes mada
,
                          to the facility in accordance with 10 CFR 50.59. The first example involved
                          operation of a safety auxiliaries cooling system (SACS) valve (1EGHV 2522E)in a
                          manner not permitted by station operating procedures. S%cifically, this valve,
                          which the UFSAR describes as providing an automatic isolation function between
                          the turbine auxillaries cooling systern and the safety-related SACS following a pipe
                          break, was failed open af ter discovery of an internal valve actuator oil leak. The
                          second example involved the placement of non safety related video camera cables
                          across multiple trains of safety related cabling, which defeated required channel
                          separation.
                          The inspectors reviewed PSE&G's March 31,1997 ietter which responded to this
                          violation. Corrective actions specified for the first issue included a procedure
                          revision which allowed operators to fail open 1EGHV 2522E (or F) under abnormal
                          circumstances. However, PSE&G engineers conducting a design basis review and
                          validation project for the service water and SACS systems later determined that the
                            10 CFR 50.59 safety evaluation used to loatify the procedure change was also
                            inadequate. Specifically, the evaluation , # 2.0 provide adequate assurance that
                          f alling open either one of the two noted au.wnatic isolation valves preserved the
                            safety function of SACS under all postulated accident conditions. The inspectors
                            reviewed the revised safety evaluation for this precedure cbenge and judged that its
                            conclusions were well founded, and appropriately focused on the safety implications
                            of the change.
                            The inspectors also assessed PSE&G's resolution of the second violation example.
                            The inspectors verified that PSE&G changed the governing procedures for
                            installation of temporary cables by including specific guidance and criteria for cable
                            routing. Additionally, focused training on this topic was provided to both fire
                                                          -
                                                                                                                _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
 
    . _ _ _ _ _ - - - _ _-                    _-_                                    .
                                                                                                                                                  .
                          .
                          .
                                                                                          18
                                              protection and radiation control personnel since these departments use temporary
                                              video cameras and cabling most frequently. No further examples of this issus have
                                              been identified.
                                        ESJ (Closed) VIO 50 354. J6125 02013:f ailure to promptly identify and correct a
                                              condition adverse to quality. This violation involved fifteen pairr of high energy line
                                              break backdraft isolation dampers which had been installed backwards in the
                                              reactor building ventilation system. This condition was self identified in 1992 as a
                                              design discrepancy, yet no action to resolve the discrepancy was taken until "re-
                                              discovered" by NRC Inspectors during a restart assessment team inspsetion during
                                              RF06. The inspectors reviewed PSE&G's November 22,1996 letter which
                                              responded to this violation, and judged tha the root cause evaluation and proposed
                                              correctiva actions were reasonable and comprehensive. PSE&G attributed the
                                              cause oi this issue to non conservative decision making, an inadequate corrective
                                              action program, and a poor appreciation for maintaining the plant in accordance
                                              with the design and licensing bases. The inspectors verified that corrective actions,
                                              which included reorientation of the reversed dampers, development of an improved
                                              corrective action program, and focused training on design and licensing bases
                                              maintenance, were implemented. PSE&G c!so performed an " extent of condition"                          c
                                              review for several other important to safety plant systems in an attempt to identify
                                              other potential design basis discrepancies. These efforts were discussed in detailin
                                              both LER 50 354/96 06and NRC Inspection Report 50 354/96 80.
                                        EQd (Closed) VIO 50 354/E96125-02023: failure to promptly identify and correct a
                                              condition adverse to quality. This violation involved an NRC inspector discovery
                                              that control rod speeds were not maintained in a manner consistent with design
l                                              basis assumptions for a startup rod withdrawal error scenario. PSE&G reported this
                                              issue to the NRC in LER 50 354/96-08 00and -01. The inspectors reviewed this
l                                              LER and PSE&G's November 22,1996 letter which responded to the violation.
r-
                                              PSE&G concluded that this issue was attributed to a failure to include UFSAR-
                                                assumed control rod speed design values as acceptance criteria in the control rod
                                                speed adjustment procedure, and that operations and engineering personnel were
    -
                                                not familiar with the associated design basis assumptions. Corrective actions
                                                included: (1) validation that all post maintenance "as-left" control rod spaed values
                                                were consistent with design basis assumptions, (2) procedure enhancements, and
                                                (3) design and licensing basis training for both operator 5 and engineers. The
                                                inspectors concluded that PSE&G's assessment and corrective actions were
                                                reasonable. The inspectors also observed that control rod speed adjustments
                                                performed during RFO7 were proper.
                                        EQJ (Closed) VIO 50-354/E96-125-03013: failure to obtain NRC approval before making
                                                  a change to the facility. This issue involved an intentional change to the operation
                                      '  _
                                                  of the station service water (SSW) pump discharge strainers, in that backwash
                                                valve operation was modified to allow a continuous (vs. cyclic) partial SSW flow
                                                  diversion from the safety auxillaries cooling system (SACS) heat exchangers. This
                                                  change was developed as a means to improve discharge strainer performance.
                                                  When implemented, the actual backwash flow diverted from SACS was much
                                                  greater than calculated (2000 gpm vs. 500 gpm). Engineering personnel then
  . . .                      .            .
                              .. . .
                                                              .
                                                                        .
                                                                                .
                                                                  .
                                                                                        .  .. .. .
                                                                                                                              ..
                                                                                                                                      . . _ _ _ _
 
    _ _ _ _ _ _ _ _ _            _-
  .
  .
                                                                    19
                            recognized that this modification invalidated the technical specification (TS)          '
                            maximum allowed ultimate heat sink (UHS) temperature limit, but argued that the
                            use of administrative controls on maximum UHS temperature were sufficient to            -
                            permit retention of the modification under 10 CFR 50.59.
                            In their November 22,1996 violation response letter, as well as during an
                            enforcement conference held on June 11,1990, PSE&G agreed that a violation
                            occurred, but did not agree with the NRC's assessment of the f acts. The inspectors
                            judged t' : since implementation of the modification rendered TS limits non-
                            conservative, it constituted a change to the facility requiring a change of TS.
                            However, PSE&G argued that the calculated vs. actual flow discrepancies qualified
                            as " discovery" following the modification, and that this discovery should have been
                            handled as a degraded condition (but was not), and characterized as " operable but    2
                            degraded"in accordance with NRC Generic Letter 91 18. As such, compensatory
                            measures such as placement of a.iministrative limits on system operating
                            parameters would have been permissible.
                            The inspectors verified that PSE&G Implemented several corrective actions +o
                            prevent recurrence of future similar issues. Specifically, PSE&G revised their 10
                            CFR 50.59 safety evaluation process to include recently issued industry and NRC
                            guidance in this area. Additionally, the operability determination process employed
                            at the station was completely revised, addmg more design and licensing basis
                            review requirements, as well as additional internal enp'neering and management
                            reviews. Extensive training for operations and engineering personnel was provided
,                            regarding maintaining the facility in accordance with design and licensing basis
'
                            assumptions. Finally, a comprehensive SACS and 60W design basis re validation
                            effort was completed in May 1997 to provide further assurance that plant operation
                            was consistent with license requirements.
                      ELD (Closed) LER 50 354/96 22: condition alone that could have prevented removal of
                            residual heat - safety auxiliaries cooling system (SACS) deficiencies. The inspectors
                            reviewed the original and two supplemental revisions to this LER. The issues
                            described in this LER were also briefly discussed in NRC Inspection Reports 50-
                            354/9610 and 97 06. Three related self-identified issues were reported: (1) SACS
                            operability could not be assured under all allowable system configurations, (2) TS
                            allowed outage times for SACS, station service water (SSW), and the emergency
                            diesel generators (EDG) were based on insufficient engineering justifications, and (3)
                            phnt operating procedures permitted a SACS configuration unsupported by the
                              system's design basis. These issues were identified during an independent SACS
                            des!gn basis review and validation effort performed as a corrective action for
                            previously-identified similar issues (see LER's 50 354/95-37,96-09, and 9615).
                            The inspectors reviewed the root cause information provided in this LER, along with
                            the stated corrective actions. PSE&G concluded that inadoquate design and
                              licensing basis reviews and validations were performed during initial system startup
                              test and operating procedure development in the mid 1980's. These inadequate
                              reviews resulted in conditions which permitted system operation outside of design
                              basis assumptions. Inadequate personnel knowledge of system licensing basis
                                      -
 
.
.
                                                20
      information contributed to the problems. When discovered, as an interim measure,
      administrative controls and compensatory actions were developed as part of a
      formal operability determination to assure that system operability was maintained.
      PSE&G corrective actions included a comprehensive SACS system flow balance,
      under all postulated operating configurations, to fully understand system flow
      characteristics. The inspectors observed portions of this testing which was
      conducted during RFO7. Additionally, the limiting conditions for operation for
      SACS, SSW, and ultimate heat sink were revised in TS amendment 100 to account
      for these and other identified design basis discrepancies such that the
      compensatory measures specified by the noted operability determination were no
      longer necessary. The inspectore verified by on site inspection that several other
      corrective actions have been completed since these issues were first identified,
      including design and licensing basis treir'ing for operations and engineering
      personnel, operating procedure revisions,10 CFR 50.59 safety evaluation process
      changes, and inservice testing program enhargements.
      The inspectors concluded that PSE&G Implemented prompt and comprehensive
      corrective actions following their self discovery of the noted issues. The stated root
      causes for these concerns were sound, and not reasonably linked to current
      performance. Since these issues were described earlier in NRC lospection Report
      50 354/97 06,in which enforcement discretion was applied not h cite a violation
      of technical specification requirements for SACS, these issues are Non-Cited
      Violations consistent with Section Vll.B.3 of the NRC Enforcement Policy. (NCV
      50 354/97 10 12)
                                          IV. Plant Support
  P2  Status of Emergency Preparedness Facilities, Equipment, and Resources
      On December 19,1997, the inspectors conducted a walkdown of PSE&G's
      Emergency Operations Facility located in Salem, New Jersey. The facility was well
      maintained and secured. Telephone systems were operational, status boards were
      cleaned of old information, and technical resource documents were available and
      up-to-date. The inspectors concluded that the facility was appropriately maintained
      in a standby condition for prompt activation.
  S4  Security and Safeguards Staff Knowledge and Performance
      On November 24,1997, security department personnel determined that guards at
      - the access control f acility failed to perform an adequate search of an Individual
      entering the protected area. The individual alarmed the metal detoctor and required
      a hands-on search prior to entry in accordance with access control requirements.
      However, according to a subsequent internal PSE&G investigation, the guards did
      not maintain positive control of the individual nor did they conduct the necessary
      search before the individual entered the protected area. This error was quickly
      recognized and the individual was returned to the access control center within
      seconds after he entered the protected area. He was properly searched before
 
    _ _ _ -    - _ _ _ _ _ _ . - _ - _ - _ _
  .
  .
                                                                                  21
                                            being permitted to re enter. According to the PSE&G investigation, inadequate
                                            communications, inattention to detail, and work site distractions contributed to this
                                            event.
                                            PSE&G performed a thorough review of this, event, and ;dentified eqveral corrective
                                            actions. Included among these actions were individual disciplinary measures, re-
                                            emphasis on maintaining positive control over individuals attempting to gain
                                            protected area access, and enhancements to security post orders with respect to
                                            personnel access control. The inspectors judged PSE&G actions following this
                                            event as thorough. This licensee identified end corrected violation of technical
'
                                            specification 6.8.1.e, which requires site security plan procedures be implemented,
                                            is being treated as a Non Cited Violation, consistent with Section Vll.B.1 of the NRC
                                            Enforcement Policy. (NCV 50 354/971013)
                                                                        V. Manaaement Mettingt
            X*                              Exit W.coting Summary
            The inspectors presented the inspection results to members of licensee management at the
            conclusion of the inspection on January 20,1998. The licensee acknowledged the
            findings presented.
            The inspectors asked the licensee whether any materials examined during the inspection
            should be considered proprietary. No proprietary information was identified.
                                                                                                                                  ii
                                                                    ~
 
.                                                                              _
  .
  .
                                                  22
                                  INSPECTION PROCEDURES USED
    IP 37551:    Onsite Engineering
    IP 61726:    Surveillance Observations
    IP 62707:    Maintenance Observations
    IP 71707: -  Plant Operations
    IP 71750:    Plant Support
    IP 90712:    Inoffice Review of Written Reports
                            ITEMS OPENED, CLOSED, AND DISCUSSED
    Qoened
    50 354/97 10-02            VIO    Failure to implement maintenance procedure.
    50 354/97 10-03            VIO    Failure to establish appropriate instructions
    50 354/97 10-04            VIO    Failure to establish appropriate procedure
    50 354/97 10-05            VIO    Insufficient corrective actions
      50 354/97-10-09            VIO    Failure to establish appropriate procedure.
?                                                                                                  ,
l
    Opened / Closed
      50 354/97-10-01            NCV Failure to implement administrative procedure
    '50 354/97-10-06              NCV Improperly performed surveillances for determining
                                        specific gravity of the 125 Vdc and 250 Vdc batteries
      50 354/97-10-07              NCV Failure to perform secondary containment isolation
                                        actuation system surveillances.
      50-354/97-10-08              NCV Removal of a RHR shutdown cooling suction line
                                        snubber.
      50-354/97 10 10              NCV Failure to establish appropriate procedure
      50 354/97-10 11              NCV Failure to perform appropriate testing
      50-354/97-10 12              NCV Failure to operate and maintain SACS in accordance
                                        with design and licensing basis
      50-354/97-10-13              NCV Failurs to perform an adequate search of an individual.
      C10D.d
      50 354/95 11                P21  Deficiency in Anchor Darling swing check valves.
      50 354/96-07 01              VIO  Failure to evaluate changes made to the facility in
                                        accordance with 10 CFR 50.59.
      50 354/E96-125 01013        VIO  Failure to plan appropriate post maintenance TS
                                        surveillance testing.
      50 354/E96-125 01023      _VID
                                _
                                        Failure to conduct appropriate post maintenance TS-  -
                                        surveillance testing.
      50 354/E96-125-02013        VIO  Failure to promptly identify and correct a condition
                                        adverse to quality.
      50 354/E96-125 02023.        VIO  Failure to promptly identify and correct a condition
                                        adverse to quality.
                                                                                                ,
 
  .
  .
                                      23
    50 354/E96125 03013 VIO Failure to obtain NRC Approval before making a change
                            to the facility.
    50 354/97 04-02
'
                        VIO HPCI system feed water injection line inoperable.
    50-354/97-07-01    VIO Failure to promptly identify inoperable electric fire water
                            pump
!
    50 354/96 22        LER Condition alone that could have prevented removal of
                            residual heat - safety auxillaries system deficiencies.
    50 354/97 27        LER Operation In a TS prohibited condition due to improperly
                            performed surveiilances for determining specific gravity
                            of the 125 Vdc and 250 Vdc batteries.
    50-354/97 28        LER TS prohibited condition failure to perform secondary
                            containment isolation actuation system surveillances.
    50 354/97 29        LER Operation in a TS prohibited condition due to removal of
                            a RHR shutduwn cooling suction line snubber.
    50-354/97 30        LER  Failed IST surveillance for reactor building to torus
                            vacuum breakers,
    50 354/97 32        LER TS required shut down due to concurrent inoperability
                            of the HPCI and RCICs
 
          - _ _ - _ _ _ _ _ _ _ _ _ - _ _ -        ____-      -.
  .
                                                                        \
  .
                                                          24
                                            LIST OF ACRONYMS USED
    nOV  Air Operated Isolation Valve
    APRM  Average Power Range Monitor
    DCP  Design Change Package
    EDG  Emergency Diesel Generator
    EPA  Electrical Protection Assemblies
    HPCI  High Pressure Coolant injection
    IA    Instrument Air
    IST  Inservice Test
    LER  Licensee Event Report
    M&TE  Measuring and Test Equipment
    NRC  Nuclear Regulatory Commission
    PClG  Primary Containment Instrument Gas
    PDR  Public Document Room
    PRNMS Power Range Neutron Monitoring System
,    PSE&G Public Service Electric and Gas
l    QA    Quality Assurance
l
'
    RBTVB Reactor Building to Torus Vacuum Breaker
    RCIC  Reactor Core isolation Cooling
    RG    Regulatory Guide
      RHR  Residual Heat Removal
      RPS  Reactor Protection System
      RWCU  Reactor Water Cleanup
      SACS  Safety Auxillaries Cooling System
      SRV-  Safety Relief Valve
      SSW  Station Service Water
      TS    Technical Specification
      UHS    Ultimate Heat Sink
                                                                  __ __
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