ML20199G750: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot change)
(StriderTol Bot change)
Line 1: Line 1:
{{Adams
#REDIRECT [[IR 05000324/1997013]]
| number = ML20199G750
| issue date = 01/23/1998
| title = Insp Repts 50-324/97-13 & 50-325/97-13 on 971109-1227. Violations Noted.Major Areas Inspected:Operations, Engineering,Maint & Plant Support.Includes Results of Maint, Engineering & FP Insps
| author name =
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
| addressee name =
| addressee affiliation =
| docket = 05000324, 05000325
| license number =
| contact person =
| document report number = 50-324-97-13, 50-325-97-13, NUDOCS 9802040338
| package number = ML20199G672
| document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 50
}}
See also: [[see also::IR 05000324/1997013]]
 
=Text=
{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ _
                          .
.
                                                  U. S. NUCLFAR REGULATORY COMMISSION
                                                                REGION 11
                                Docket Nos:        50-325, 50 324
                                license Nos:        DPR 71. DPR 62
                                Report No:          50-325/97-13. 50-324/97 13
                                Licensee:          Carolina Power & Light (CP&L)
                                Facility:            Brunswick Steam Electric Plant, Units 1 & 2
                                Location:            8470 River Road SE
                                                      Southport, NC 28461
                                Dates:              November 9 - December 27, 1997
                                  Inspectors:        C. Patterson Senior Resident Inspector
                                                      E. Brown Resident inspector
                                                      G. Guthrie, inspector in Training
                                                      J. Coley Reactor inspector (M1.3. M8.6)
                                                      J. Lendhan. Reactor Inspector (E1.1. E1.4. E5.1. E8.3.
                                                        E8.4. E8.5)
                                                      C. Doutt. Senior Instrumentation and Controls
                                                        Engineer. Office of Nuclear Reactor Regulation
                                                        (E1.1. E1.2. El.3)
                                                      G. Wiseman. Reactor Inspection (F2.1. F2.2. F2.3,
                                                        F3.1. F5.1 F6.1. F7.1)
                                  Approved by:        M. Shymlock. Chief. Projects Branch 4
                                                      Division of Reactor Projects
                            9802040330 900123
                            PDR
                            G    ADOCK 05000324
                                              PDR
                                                                                                  Enclosure 2
                                                                                                        _ _ _ _ _ _ _ _ _ _ _ .
 
                                                                                    i
  *
                                                                                    \
                                                                                    1
.
                                      EXECUTIVE SUMMARY
                        Brunswick Steam Electric Plant. Units 1 & 2
                    NRC Inspection Report 50 325/97 13. 50-324/97-13
    This integrated inspection included aspects of licensee operations,
    engineering, maintenance, and plant support. The report covers a 6-week
    period of resident inspection; in addition. It includes the resu'ts of
    maintenance, engineering, and fire protection ir,pections by regional and
    headquarters inspectors.
    Operations
    e    The inspector concluded that u.e cold weather program has been
          satisfactorily implemented. Adequate contingency plans and operator
          checks for proper operation of the systems were noted in the procedures.
          Section 01.1).
    *    The inspector concluded. from a safety system walkdown, that the
          Containment Atmospheric Dilution system was being maintained as designed
          (Section 02.1).
    *    The clearance reviewed was prepared. authorized, and implemented in
          accordance with procedure (Section 02.2),
    e    The inspector concluded that the Plant Nuclear Safety Committee meeting
          provided an effective review of Unit I readiness for restart (Section
          07.1).
    e      Inspe.; tor review determined that clearance records were not retained in
          accorcance with Technical Specifications (TS). The failure to maintain
          clearance records in accordance with TS was a violation (Section 07.2).
    *    The control of a short duration mid-cycle o:tage was excellent (Section
          07.3).
    *    Licensee investigation determined that removal of the IB Reactor
          feedwater Pump at too high a power level caused larger than expected
          level transients. These transients combined with the improper
          functioning of the level contacts in the Reactor Recirculation Run back
          logic circuitry, resulted in the November 5-6. 199/ run backs (Section
          08.3).
    *    The inspector concluded that the licensee's control of the 2C and 20
          electrical bus maintenance was weak because they did not recognize DG in
          oberabilityconditionsduringtheimplementationoftt.eirclearance
          ( ection 08.4).
 
  F *
                                                                                      1
                                                2
      Maintenance
      e    Movement of the spent fuel shi) ping cask was perforrxo in accordance
            with methodology approved by t1e NRC in a letter dated December 2, 1997.
            Adequate supervisory oversight was present during movement of the cask
            (Section M1.1).
      *    The inspector observed performance of calibration of two Reactor Core
            Isolation Cooling (RCIC) pressure switches. The work activities were
            completed without any identified questions or concerns (Section M1.2).
      *    Maintenance activities observed relating to equipmert qualification of
            electrical equipment were found to be conducted in a thorough and
            effective manner (Section M1.3).
      .    A violation was identified for a preventive maintenance procedure not
            indicating specific E0 requirements. This omission resulted in
            deficient Nelson flame seals in motor control centers not being detected
            during scheduled preventive maintenance activities (Section M1.3).
      *    The licensee continues to struggle with proper dispositioning of
            abnormal indications. The failure to maintain the Daily Surveillance
            Report in accordance with procedure was a violation. Abnormal values
            observed fer the Steam Jet Air Ejector radiation monitor and subsequent
            test indicated potential fuel failure for Unit 1 (Section M3.1),
      *    The licensee identified that the Unit 2 Core Spiay sparger differential
            alarm setpoints were outside of the TS allowable range. The cauce was
            attributed to voiding of the sparger nozzles similar to the phenomenon
            identified previously on Unit 1. The alarm setpoints were adjusted and
            the associated documentation was updated (Section M8.5).
      -Engineerino                                                                    >
      +    An additional example of a violation was identified for an inadequate
            procedure for the conduct of E0 maintenance (Section E1.4).    Two
            inspector followup items were identified to review revisions to
            instrument setpoint procedures and to review terminal block leakage
            current evaluations (Section El.1 and Section E1.4).
      *    A weakness was identified regarding a procedure reference to a drawing
            for accident temperature data which was not available for use and
            wording inconsistencies in the procedure (Section E1.1).
      *    The licensee was making progress in resolution of the technical issues
            and closure of CRs and JCOs (Section E1.4). The licensee training and
            qualification for E0 personnel meets NRC requirements (Section E5.1).
              Instrument setpoint calculations were technica ly adequate and complied
            with NRC requirements (Section E1.2).
i
 
                                                            _ _ _ _ _ _ _ _ _-
      .
    .
                                                  3
        Plant Support
        .      The ins)ector determined that each of the locked high radiation area
                doors w11ch were checked were locked.    lhe ins)ector concluded that the
                licensee is satisfactorily controlling locked ligh radiation areas in
                the plant (Section Rl.1).
        .
                The inspector determined that several poor radiological work practices
                existed in a radioactive material storage area (Section Rl.2).
        *
                The inspector found the status and condition of the protected area fence
i                to be satisfactory (Section S2.1).
        .
                Corrective maintenance on degraded fire protection systems was
i
!
                accomplished in a timely manner. The maintenance and material condition
                of the fire protection equipment and features were satisfactory
                (Section F1.1).
        .      The inspector concluded that silicone foam penetration seal field
                verificction documentation was maintained by the licensee. The
                inst 311ation and repair procedures for penetration seals provided
                adequate guidance to ensure that materials were installed per design
                requirements. However, the designs were not supported by seal testing
                documentation, vendor data and inspection criteria, installer
                qualification and training records, and engineering evaluations that
                satisfy the guidance of Generic Letter 86-10 for deviations from the
                fire barrier configuration qualified by tests (Section F2.2).
        .
                The inspector concluded that fire door surveillance procedures and
                acceptance criteria for verification of fire door clearances were in
                accordance with National Fire Protection Association (NFPA) guidance.
                However, an updated Final Safety Analysis Report (UFSAR) discrepancy
                associated documentation of fire door and frame evaluations was
                  identified (Section F2.3).
        .      General housekeeping was satisfactory. Fire retardant plast.ic sheating
                and film materials were being used. Lubricants and oils were properly
                stored in approved safety containers. Controls for combustible gas bulk
                storage and cutting and welding operations were being enforced.
                Controls were being properly maintained for limiting transient
                combustibles in designated separation zones and other restricted plant
                areas (Section F3.1).
        .
                The fire brigade organization and qualification training .act the
                requirements of the site Procedures. Fire brigade turnout gear and fire
                  fighting equipment were being properly maintained (Section F5.1).
        .
                The coordination and oversight of the tacility's fire protection program
                had been reassigned from the previous Loss Prevention Unit organization
                to shift. Operations. The new organizat.onal structure met NRC
                guidelines and the licensee's fire protection program requirements
                  (Section F6.1).                                                          .
                                                                                          l
  1
9
      .
              -
 
                                                    ..-  -      -      -. .
                                    4
. The 1997 Nuclear Assessment Section assessment of the facility's fire
  protection program was comprehensive and was effective in identifying
  fire protection program performance deficiencies to management. Planned
  corrective actions in response tc the audit issues were substantial and
  included a fire p.'otection reorganization (Section F7.1).
 
                                    _ _ _ _ _ _ - _ _ _
  .
                                                        ReDort Details
    ~ Summary of Plant Status
    Unit I returned to power o)eration on November 14. 1997, following a mid-cycle
    outage that began on Novem)er 5. 1997, to remove leaking fuel assemblies. Two
    leaking fsel assemblies were identified and removed during the mid cycle
    outage.    However, indications of a potential fuel leaker remained after the
    unit returned to full power operation. At the end of the report period the
    unit had been on-line 42 days.
    Unit 2 operated continuously during this report period. At the end of the
    report period the unit had been on-line continuously for 59 days.
    Due to concerns about the control room dose, the licensee imposed an
    administrative limit on lodine until a Technical Specification (TS) amendment
    submitted was a) proved.    The licensee made a orocedure change to
    Administrative procedure 0Al-81. Water Chemistry Guidelines, setting the limit
    at 0.1 microcurie per gram dose equivalent L 'ine 131 compared to the TS value
    of 0.2 microcurie per gram. Also, the licet ;e has been providing weekly
    water chemistry data to NRR and the Resident Inspector for review. None of
    the data reviewed has exceeded the administrative limit.
    Due to a reconstitution of the Environmental Qualification (EO) program and
    items identified, there are 12 of 24 Justification for Continued Operation
    (JCO) that remain open for both units. The following provides the status of
    the EQ JCOs and associated Engineering Service Requests (ESRs):
I
            Closed
.
'
!
              1)  ESR 97-00087. E0-Type JC0 for Improperly Configured Conduit Seal.
              2)  ESR 97-00574 Greyboot Connectors.
              3)  ESR 97-00329 (old ESR 96-00625). E0 Type JC0 for EQ Fuses Without
                  a Qualification Data Package (00P).
              4)  ESR 97-00289. Post A cident Sampling System (PASS) Valve Limit
                  Switch Panel Wiring.
              5)  ESR 97-00238. JC0 for Standby Gas Treatment Motor Operated Valve
                  (MOV) Position Indicator Rheostat.
              6)  ESR-97-00534. GE c'                    Type Terminal Strips.
              7)  ESR 97-00513. In-b                      Drywell Electrical Penetrations.
              8)  ESR 97-00535. Target Rock Solenoids TB Spray.
              9)  ESR 97-00449, Degraded Junction Boxes.
            19)  ESR 97-00250. Conduit Union in EQ Boundary.
            11)  ESR 96-00425. Evaluation of E0 sealants.
            12)  ESR 97-00523. High Pressure Coolant Injection Auxiliary Oil Pump
                  Motor Unit 1.
            0P10
            13)  ESR 97-00446. GE Radiation Detectors. closure date to be
                  determined (TBD).
            14)  ESR 96-00503. Associated Circuit E0. closure date TBD.
                                                                                            . _ _ _ _
 
                                      _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
                                                                                        ..
  .
                                                                                2
          15)    ESR 97-00330 (ola ESR 96 00501). Motor Control Center (MCC) E0 was
                closed by the licensee, but was reopened - closure date TBD.
          16)    ESR 96-00426. Evaluation Quality class and E0 classification of
                PASS valves was scheduled for completion June 6, 1997. but closure
                date is TBD.
          17)    ESR 97-00529. Failure of Unit 1 Drywell Motor, closure date TBD.
          18)    ESR 96 00587 PASS Valves, closure date TBD.
          19)    ESR 96 00627 ODP for Marathon 300 Terminal Blocks was scheduled
                for completion December 31, 1937 but revised to August 1. 1997,
                but closure date is now TBD.
          20)    ESR 97-00229. JC0 for GE Condition Report (CR) 151 B Terminal
                Blocks was scheduled to be completed September 1, 1997, but
                closure date is now TBD.
          21)    ESR 97-00256. Main Steam Insulation Valve Hiller Aci . tor JCO. was      -
                scheduled for completion September 2, 1997. but closure date is
                now TBD.
          22)    ESR 97-00343. Qualification of Kulka Model 600 Terminal Blocks was
                scheduled for completion September 1. 1997, but closure date is
                now TBD.
          23)    ESR 97-00435. MCC Fittings, closure date TBD.
          24)    ESR 97-00602. Solenoid Valve Field Wiring, closure date TBD.
.
'
    In summary Unit I returned to power operation following completion of a mid-
    cycle outage.    Unit 2 o)erated continuously; however there were 12
    outstanding JCOs in the E0 area for both units.
                                            I. Ooerations
    01    Conduct of Operaticns
    01.1 Cold Weather Preparation
    a.    Insoection Scone (71714)
          The inspector reviewed the licensee's cold weather program to determine
          whether it had been effectively implemented.
    b.    Observations and Findinas
          The inspector reviewed the licensee's cold weather 3rogram for adequacy
          and implementation by reviewing their Cold Weather 3111 and Freeze
          Protection Procedure. Operating Instruction 001-01.02: Fire Protection
          Procedure 0FPP-024. Freeze Protection of Fire Suppression System; and
          Preventive Maintenance Procedure OPM-HT001. Preventive Maintenance on
          Plant Freeze Protection and Heat Tracing. The inspector determined that
          the procedures were adequately implemented. Additionally, the
          procedures were adequately employed on multiple cold weather days. as
          observed by the inspector.
          The inspector conducted a walkdown of plant syn. , which were exposed
          to cold weather. Systems which were heat traced were observed for
          adequacy. The inspector looked for systems that did not have cold
<
                                                                                  ..
                                                                                      .      . __ .. .
                                                                                                      _
 
                            _    . _ _ _ _ _ _
                                                3
        weather. heat trace installed. The inspector determined that the
        operation of the Makeup Water Tank system heat trace was not controlled
        by any procedure. The licensee stated that this heat trace system was
        being controlled b," operator knowledge only. The licensee initiated a
        procedure change request to place this heat trace system into their cold
        weather procedures. The inspector noted on the Unit 2 Condensate
        Storage Tank. High Pressure Coolant Injection (HPCI)/ Reactor Core
        Isolation ~ Cooling (RCIC) level switch vent line that a six inch portion
        of the lagging was missing at the top of the vent line and that the tin
        shielding was missing around the lagging at an elbow on the vent line.
        The lagging was wetted and degraded at the elbow. The inspector
        discussed these two items with the licensee. The licensee did not
        warrant these deficiencies as requiring corrective action. The
        inspector did not find other systems requiring heat trace that were not
        heat traced based on present system conditions and projected use of the
        systems observed.
  c.  Conclusions
        The inspector concluded that the cold weather program has been
        satisfactorily implemented. Adequate contingency plans and operator
        checks for proper operation of the systems were noted in the procedures.
  02    Operational Status of Facilities and Equipment
  02.1 Containment Atmosoheric Dilution (CAD) System Walkdown
.
'
  a.  Insoection Scope (71707)
        On December 10. 1997, the inspector performed a walkdown of the CAD
        system in the Nitrogen and Off-Gas Services Building.
  b.  Observations and Findinos
        The CAD system is described in Updated Final Safety Analysis Report
        (UFSAR) Section 6.2.5. Combustible Gas Control in Containment. The CAD
        system provides long-term nitrogen makeup after a Loss of Coolant
        Accident (LOCA).    This function is accomplished by vaporizing liquid
        nitrogen and feeding it into containment as required to maintain an
        oxygen concentration at or below five percent. The system is designed
        to Engineered Safety Feature (ESF) standards, all equipment for CAD
        service is designed with suitable redundancy and interconnections such
        that no single failure of an active component will render the system
        inoperable. This equipment includes one liquid nitrogen storage vessel.
        two electric vaporizers, two flow-regulating stations. flow and
        temperature indicators. and appropriate redundant valves and
        interconnecting piping.
        The inspector traced the system piping in the Nitrogen and Off-Gas
        Services Building. The configuration was compared to plant drawing
        0 02560. Containment Atmospheric Control System. The configuration was
        found to be like the plant drawing. The inspector observed an inch of
            -
                      -
                                                              -                  ,
 
                                          _ _ _ _ _ - _ _ _ _ _ - _ _ _
                                          4
          frost on the outside of the piping insulation on both sides of valve
        HV-11.  This valve is a manual isolation between the nitrogen tank and
        an 85 pound pressure regulating valve.
        The inspector questioned why the frost was on the line. The licensee
        stated that the 90 pound relief valve setpoint was near the controlling
        pressure of the 85 pound regulator and some nitrogen was venting off.
        The redundant pressure regulating valve was isolated and it's isolation
        valve (HV-12) was closed. The inspector questioned by keeping HV-12
        closed, if the system was single failure proof. The licensee initiated
        CR 97-04128. CAD Tank Isolation Valve, to address this issue, The
          licensee concluded that no automatic action was required to address a
        LOCA.  Manual alignment of the pressure regulator was acceptable since
        this was a long term post-LOCA action.
    c.  Conclusions
        The inspector concluded, from a safety system walkdown, that the CAD
          system was being maintained as designed.
    02.2 Clearance Verification
l    a.  Insoection Scoce (71707)
        The inspector reviewed the tagout for the Unit 2 Residual Heat Removal
                        -
          (RHR) system to verify proper clearance preparation, authori7    n. and
          implementation,
      b. Observations and Findinas
          On December 10. 1997, the inspector performed verification of the proper
          alignment and tagging of clearance 2-97-1781 on the Unit 2 RHR System.
          All accessible components were verified to-be in the proper position
        with the appropriate tags in place. The inspector reviewed Nuclear
          Generation Group Standard Procedure OPS-NGGC-1301. Equipment Clearance.
          The clearance package was adequately prepared, authorized, aad
          implemer.ted. The inspector subsequently verified proper clearance
          removal for those accessible components.
      c.  Conclusions
          The clearance reviewed was prepared, authorized, and implemented in
          accordance with procedure,
i
  w
 
                                    . _ . _ _ _ _ _ _
                                                                              . .  . ..  ._ .
                                                                                                  .
                                                          5
  07    Quality Assurarm in Operations
  07.1 Restart Plant Nuclear Safety Committee (PNSC)
    a.  Insoection Scone (71707)
        On November 11 and 12. 1997, the inspector attended the Unit 1 PNSC
        restart assessment following a mid-cycle outage to replace two leaking
        fuel assemblies,
    b.  Observations and Findinos
        On November 11, 1997. PNSC was convened to review Unit I readiness for
        restart. The committee reviewed the fuel sipping results and core
        reload.  Other maintenance activities during the outage were also
        reviewed.
        The meeting was conducted in accordance with TS with attendance by all
        primary members, with no alternates. The meeting provided a thorough
        discussion of all agenda items. The PNSC Chairman concluded that the
        discussion of recirculation pump runbacks that occurred on November 5.
        1997, during removal of the reactor feed pumps during the planned
        shutdown was not complete.                    This item was statused as a restart
        constraint requiring another PNSC review prior to restart. Noteworthy
        in the review was the risk assessment review conducted for a failed
        Control Rod Drive (CRD) pump. During the mid-cycle outage one of the
        two CRD pump motors failed. The Probabilistic Safety Analysis (PSA)
        person attended the comnittee meeting and presented the results from
        running the risk assessment model considering failure of both CRD Jumps.
        This risk was determined acceptable based on other TS required higi
        pressure injection sources such as HPCI and RCIC.
        On November 12. 1997, the inspector attended a second meeting. In this
        meeting discussion was held regarding the problem with run backs and it
        was concluded that this was due to a design deficiency that was already
        corrected and installed on Unit 2 and scheduled for Unit 1 at the time
        of the next refueling outage,
    c.  Conclusions
        The inspector concluded that the PNSC meeting provided an effective
        -review of Unit I readiness for restart.
  07.2 Retention of Clearance Records
    a.  Insoection Scope (71707)
\
        The inspector reviewed whether configuration management documents,
        specifically ciearances, were retained in accordance with TS 6.10. This
        specification requires that facility records be retained in accordance
        with the American National Standards Institute (ANSI) N45.2.9-1974
        Collection. Storage, and Maintenance of Quality Assurance Records.
        -                                                                                      _
 
                  _ _ _ _ _ _ _ - _ - _ _ _    _ _ _ _      __  --
                                                                            -
                                                        6
b.  Observations and Findinas
      During ins)ector review of clearance errors which resulted in damage to
      the Unit 23 recirculation pump seals, the licensee was unable to locate
      a clearance hung to facilitate repairs on the recirculation motor oil-              -
      cooler. Tha clearance. 2-97-1531. was hung _resulting in a configuration
      change for the B recirculation pump, but no maintenance on the system
      was performed. The clearance was removed from the field, thus restoring
      the system, and " rolled back" to allow use at a later date.
      Subsequently, a scheduler requested the clearance be deleted due to the
      repair activities being complete and approved without need for the
      clearance boundary. As a result of the deletion of the clearance, no
      record of the change in plant configuration was retained.
      The inspectoi : viewed TS 6.10. UFSAR Section 1.8. Regulatory Guide
      1,88, and ANS1 N45.2.9-1974.                      fhe inspector questioned the correctness
      of not retaining the clearance. Since a configuration change did occur
      despite the recirculation motor cooler activities not needing the cooler
      isolated. Nuclear Records Management Procedure ORMP-001. Indexing of
      Plant Records. defined those records required to be retained to satisfy
      the 0A requirements stated in ANSI N45.2.9-1974. Discussion with the
      licensee revealed that the records required to be retained did not
      include clearances. The inspector reviewed the Nuclear Generation Grou)
      Standard Procedure OPS-NGGC-1301. Equipment Clearance, and the Brunswicc
      Required Records List.                Neither document required that clearances be
      retained.
      TS 6.10 requires facility records shall be maintained in accordance with
      ANSI N45.2.9-1974. ANSI N45.2.9-1974, in Section 3.2.7. Retention of
      Records. states that Appendix A to the standard defined the types of 0A
      records and the recommended retention periods. The failure to maintain
      data sheets or logs on equipment alignment consistent with ANSI N45.2.9-
      1974 is a violation. This violation is identified as VIO 50-325
      (324)/97-13-01. Failure to Retain TS Required-0A Record.
  c.  Conclusion
      Inspector review determined that clearance records were not retained in
      accordance with TS. The failure to maintain clearance records in
      accordance with TS was a violation.
07.3 Mid-Cycle Outaae (71707)
  a.  Insoection Scope
      The inspector reviewed the mid-cycle outage activities to remove the
      leaking fuel assemblies.
  b.  Observations and Findinas
      Unit 1 was returned to power operation on November 14. 1997. This
      completed a mid-cycle outage in eight days. The unit was shutdown.
                                                                                                _-
 
                                  _ _ _ _ _ _ _ _ _ _ ___          _    __
  .
                                                          7
          leaking fuel assemblies identified, removed, fuel reloaded and returned
          to power o)eration. This short duration outage was the quickest on
          record. T11s was accomplished with plant personnel without any major
          problems. This outage was planned and controlled similar to a regular
          refueling outage.
      c. Conclusions
          The control of a short duration mid-cycle outage was excellent.
    08    Miscellaneous Operations Issues (92700, 92901)
    08.1 (Closed) Unresolved Item (URI) 50-325/96-15-01:                      Vessel Disassembly
          Without Secondary Containment.
          During a refueling outage, the reactor vessel head and steam
          dryer /separatorr assemblies were removed from the reactor vessel without
          secondary containment integrity (SCI) established. This issue was
          reviewed by the NRC Office of Nuclear Reactor Regulation. It was
          determined that the removal of the nead and assemblies without SCI
          established were not activities prohibited by TS 3.6.5.1. The potential
!        for load handling accidents was a safety cuestion that has been reviewed
!
'
          by the NRC. However, maintenance of SCI curing vessel disassably was a
          logical extension of the defense-in-depth ap3 roach used in addressing
          the heavy loads issue and encouraged by the 4RC. The licensee's action
          in proceeding with vessel disassembly was not conservative. The
          licensee implemented controls during the Unit 2 refueling outage to
          maintain secondary containment operable during vessel disassembly. This
          issue was thoroughly evaluated as part of the licensee's Safe Shutdown
          Risk Management Assessment.
    08.2 (Closed) Violation V10 50-325(324)/97-02-01:                        Locked Valve Out of
          Position
          The licensee's response to this violation was dated May 5, 1997, and was
          accepted by the NRC in a letter dated May 23. 1997. The corrective
          actions described in the response letter were verified as complete by
          the inspector. This violation is closed.
    08.3 (Closed) URI 50-325/97-12-03:                    Recirculation Pumo Run backs
          On November 5. 1997, the licensee began a c0ntrolled shutdown for the
          Unit 1 forced outage in order to replace leaning fuel bundles. During
          the shutdown. Unit I received two recirculation pump run backs to the 45
          percent limiter. During the second run back the five percent buffer
          region was entered and exited in accordance with procedures.
)          Subsequently. no other transients or run backs were ercountered while
          removing the Reactor Feedwater Pumps (RFPs) from service. The licensee
          preliminarily attributed the first run back to a malfunction of the 1B
          discharge check valve causing diversion of the 1A RFP through the 1B
          discharge valve to the main condenser. The final analysis was provided
          in the root cause analysis for CR 97-3917. Unit 1 Plant Transients While
 
    -.    . _ _ _ - - _ ..
                                - - - - - - - - - = - - - -
                                                            8
        Removing a Reactor Feed Pump from Service. The inspector reviewed the
        analysis and noted that the root cause attributed the run backs to the
        removal of the RFPs at too high of a power level and a design problem in
        the a) plication of the Metal-On-Silicon Field Effect Transistor (MOSFET)
        switcl. The MOSFET was used in the 45 percent recirculation pump run
        back logic to indicate the below 182 inches reacter water level contact
        which is one of two contacts required to initiate the run back.
        Reactor water level perturbations are expected during the removal of the
        RFPs from service: however the magnitude of these perturbations seen for
        these events were outside of the operators expectations. The root cause
        analysis stated that removal of the RFP at 65 percent power was
        inappropriate in that 65 percent during this evolution has changed since
        power uprate. Before power uprate. RFPs were removed from service 3er
        10P-32, Condensate and Feedwater System Operating Procedure, at or )elow
        65 percent. Under current conditions 65 percent is approximately
        equivalent to 68 percent power pre-uprated power. The analysis
        attributed the magnitude of the perturbations to removal at too high of
        a power level. In addition, the licensee determined that when the first
        RFP was taken out of service, the less than 20 percent RFP flow contact
        for the 18 pump was made up and with the MOSFET improperly indicating
        below 182 inches water level the run backs were received. The design of
        the MOSFET causes the contact to not be able to properly position itself
        u'aon loss of the constant voltage supply. Therefore interruptions in
        tle voltage will cause the MOSFET contact to not function as designed.
        The second Run back was also attributed to the MOSFET.                    The licensee
        intends to replace the MOSFETs in the next Unit 1 outage, The inspector
        noted that the MOSFETs had already been replaced in Unit 2.
        The licensee is reviewing plant operation to determine the appropriate
        power level for removal of the RFPs from service. Based on licensee
        satisfactory comaletion of the investigation into the cause for the
        multiple run bac(s on November 5-6, 1997 this item is closed.
  08.4 (Closed) URI 50-325(324)/97-12-04:                    Diesel Doeration Low Voltace Auto
        Start Defeated
        The inspector reviewed the licensee's root cause investigation CR 97-
        03683, 4KV Bus 2C/2D Clearances. The licensee's investigation
        determined that the number 3 diesel generator (DG) undervoltage relay
        had been disabled in the same manrer as the number 4 DG during similar
        maintenance activities on different days.
          The inspector verified that the licensee did not exceed TS action,
          limiting condition for operation, or time requirements for both
          electrical bus maintenance activities. The inspector found that, on
          October 9. 1997, the plant was under a TS action statement requirement
          per TS 3.8.2.1. to restore the inoperable bus to operable within 8
          hours, or be in hot shutdown within the next 12 hours. The electrical
..
          bus was not restored, in this case, for 12 hours and 58 minutes. This
          plant condition was not recognized as a problem until the root cause
          investigation was performed. The root cause investigation was found to
                                                                            _    _            __
 
  _                                  _ _ - _ _ _ _  _ _ _ _ _ - _ _ _ _ _ _ _ _
                                                  9
          be adequate.  The ins)ector concluded that the licensee *s control of the
          2C and 2D electrical aus maintenance was weak because they did not
          recognize that the DG would be inoperable during the implementation of
          their clearance. This item is closed.
                                          II. Maintenance
    M1    Conduct of Maintenance
    M1.1 Spent Fuel Cask Movement
      a. Inspection Scooe (62707)
          The inspector observed transfer of the spent fuel shipaing cask from the.
          117 foot elevation to the transport v'hicle and from t1e transfer
          vehicle to the 117 foot elevation of the Unit 1 Reactor Building.              s
      b. Observations and Findinas
          On December 8. 1997, the inspector observed the removal of the spent
          fuel shipping cask, with fuel in the cask from the 117 foot to the 20
          foot elevation in the Unit 1 Reactor Building. On December 15, 1997,
          the inspector observed shipping cask movement, without fuel in the cask,
          from the 20 foot elevation to the 117 foot elevation in the Unit 1
          Reactor Building. During both evolutions the cask was transferred with
          the valve box covers removed while being moved by the non-single failure
          proof yoke. Approval for use of a non-single failure proof yoke for
          movement of the cask with the valve covers removed was granted to the-
l        licensee by the NRC in a letter dated December 2, 1997. Upon reaching
          the transfer vehicle on December 8. 1997. the cask was wiped down to
          reduce contami.1ation levels. During both movements the inspector noted
          that the area was adequately posted for the radiological conditions
I        present and i ealth pnysics personnel were present. The inspector noted
          that adequate maintenance supervisory oversight was present for both
          cask movements.
          Subsequent surveys of the cask after removal from the Reactor Building
          revealed that the shipment exceeded required limits. This event was
          captured in CR 97-4161. S)ent Fuel Cask (IF-300). The cask was returned
          to the Reactor Building w1ere additional decontamination was conducted.
          The licensee attributed the contamination levels seen to leaching of the
          contamination due to changing temperatures and weather conditions.
      c. Conclusions
          Movement of the spent fuel shi) ping cask was performed in accordance
          with methodology approved by t1e NRC in a letter dated December 2. 1997.
          Adequate supervisory oversight was present during movement of the cask.
                                                                                    -_ _a
 
                                                    _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
  ,
                                            10
    M1.2 RCIC Turbine Exhaust Diaphraam High Pressure Instrument Calibration
      a.  Insoection Scoce (61726)
          The inspector observed the performance of Maintenance Surveillance Test
          2MST-RCIC230. RCIC Turbine Fxhaust Giaphragm High Pressure Instrument
          Channel Calibration, for the pressure switches 2-E51-PSH-N012A and 2-
          E51-PSH-N012C.
      b.  Observations and Findinas
          On December 24. 1997, with Unit 2 at 100 percent power the inspector
          observed the channel calibration for RCIC pressure switches 2 E51-PSH-
          N012A and 2-E51-PSH-N012C.    The inspector verified that duriug the
          performance of this channel calibration that HPCI and Automatic
          Depressurization System (ADS) were o)erable and that no othar work
          activities were being conducted whic1 could cause an inadvertent
          isolation. This test verified that, upon sensing of a high pressure
          condition between the t'arbine exhaust dia)hragms, an isolation signal is
          sent in accordance with TS 4.3.2.1 and Ta)les 3.3.2-2(4.b.6) and 4.3.2-
          1(4.b.6)
          The inspector reviewed the work request / job order (WR/J0) AKNU 19 and
          the governing procedure 2MST-RCIC230. The procedure in use was verified
          to be the correct revision and the test instrumentation in use was
          within the allowable calibration duration. The inspector observed the
'        procedure in use at all work locations and adequate communication was
          maintained throughout the test. The work observed was completed
          satisfactorily with no observed concerns.
    c.  Conclusions
          The inspector observed performance of cal:uration of two RCIC pressure
          switches. The work activities were completed without any identified
          questions or concerns.
    M1.3 General Comments
      a.  Insoection Scone (62700)
          The inspector examined the following work activities involving EQ
          electrical equipment to verify maintenance implementation of EQ
          requirements.
          *
                  WR/JO 97-ALVT-002 Verified Calibration of Unit 1 Loop B Residual
                  Heat Removal (RHR) Service Water Pressure Switches Tag No.
                  1-SW-PS-1176 B and 1-SW-PS-11760
          *
                  WR/JO 97-AGDR-002 Verified Calibration of Unit 1. Loop A. RHR Flow
                  Transmitter (1-E11-FT-N015A). Converter (1-E11-FY-5119A). Square
                  Root Converter (1-E11-FY-K600A)
                                                                                              - ___
 
                                                _____ _ ____
                                          11
        .
                WR/JO 97-AAAS-002 Unit 2. Loop B. RHR Breaker Test in compartment
                DM 5 of GE IC 7700 Series MCC 2XB-2 Division Il
    b.  Observations and findinos
        The above work was ,m cformed with the work packages present and in
        active use. Technicians were skillful, experienced, and knowledgeable
        of their assigned tasks. However, on December 10, 1997, while observing
        Instrumentation and Control (I&C) maintenance personnel perform work
        activities in accordance with WR/JO 97-AAAS-002, the inspector noted
        that one of the multiple cable electrical penetrations in the top of MCC
        2-2XB-2 did not have Nelson flame guard putty on the inside surface as
        required by Maintenance Procedure OMMM 016. Environmental Qualification
        Maintenance Program. Revision 4. to properly seal the penetration. The
        inspector examined the putty installation on the top of the MCC cabinet
        for each of the penetrations and found the putty seal severely damaged
        on a second multiple cable penetration. In addition, cables were loose
        in both of the multiple cable penetrations. The applicable
        Environmental Otalification Data Package (ODP). ODP 67, requires missing
        or disturbed Nelson putty seals to be repaired or replaced. However,
        the PM procedure used to maintain and inspect the MCC's (PM Procedure
        OPM-MCC002. Revision 7. PM of GE Motor Control Centers and Switchboards)
        did not have inspection requirements or acceptance criteria to ensure
        that putty seals were properiy sealing the cabinets. On September 17.
f      1997, a three-year PM conducted on MCC 2-2XB-2 would have identified
l      this discrepancy had procedure OPM-MCC002 included the acceptance
        criteria for the Nelson flame seal putty. A subsequent inspection
        performed on December 11. 1997 by the licensee, of 22 MCCs found an
        additional three MCC cabinet penetrations with damaged Nelson putty
        seals. In addition. 15 3ercent of the cables inspected in cabinet
        penetrations had putty w1ich appeared not to fully adhere to the cable
        in some areas. Failure of the procedure to implement E0 requirements
        for Nelson autty seals is identified as VIO 50-325(324)/97-13-02.
        Inadequate 3rocedure for the Conduct of E0 Preventive Maintenance.
    c. Conclusions
        Maintenance activities observed related to E0 of electrical equiament
        were found to be conducted in a thorough and effective manner, iowever,
        a violation was identified for a PM procedure not indicating specific E0
        requirements. This omission resulted in deficient Nelson flame seals in
        MCCs not being dettcted during scheduled PM activities.
  M3    Maintenance Procedures and Documentation
  M3.1 Steam Jet Air Eiector Off-Gas Radiation Monitor increase
    a.  Inspection Scoce (61726)
        The inspector reviewed selected sections of Operating Instruction 101-
        03.1. Control Operator Daily Surveillance Report to ensure that
                                                                                  i
                                                                                  1
                                                                                  ,
            e
 
                                      _______
  .
,
                                                12
      appropriate and prompt actions were taken to address abnormal TS
      surveillance values,
    b. Observations and Findinos
      On December 2. 1997. Unit 1 was in mode 1 at 100 percent power.          The
      inspector reviewed the daily surveillance report as contained in
      Attachment 1 to 101-03.1 for November 30 through December 1. 1997.          The
      inspector noted that the values for the Steam Jet Air Ejector (SJAE)
i
      off-9as radiation monitors on aage 26 were between 1570 and 1780
      millirem per hour (mR/hr) whici was greater than the T3/ Operating Limit
;      value of 1000 mR/hr. The SJAE off-gas radiation monitors provide for
      the detection of fuel element failures. The radiation levels are
      recorded in 101-03.1 to provide an indication whether SJAE off-gas
      radiation levels are approaching the alarm setpoint, which serves to
      ensure that dose rates for gaseous effluents do not exceed the limits
l
      prescribed in TS 3.11.2.1. Dose Rate.
l      The inspector reviewed the associated procedures, work tickets, and
      discussed the abnormal values with the licensee. Step 4.2 c 'f 001-03.1
      required the control operator to red circle all values wt            are not
      within required limits. The inspector noted no indication on the
      attachment or in the operator logs that action had been taken or was
      expected to be performed to address the out-of-range values. Subsequent
      reviews of the daily log entries by the inspector indicated continual
      abnormal values and no red circles.        These failures were recorded in CR
      97-4136. Daily Surveillance Report.        The failure to red circle values
      not within required limits is a violation.        This violation is identified
      as VIO 50-325/97 13-03. Failure to Note Abnormal TS Surveillance Values.
      CR 97-4100. Questioned OG Data / Fuel Leak      indicated that on December 3,
      1997, a step increase of approximately 200 mR/hr was seen on the
      radiation monitor Subsequent sample results have shown an increase in
      the Sum of Six value ano changes in the fuel reliability index which are
      signs of potential fuel failure. In addition, the inspector noted that
        incorrect sensitivities were used during the November 25, 1997.
      adjustment of the SJAE radiation monitor alarm setpoilts. This was
      documented by the licensee in CR 97-4046. SJAE Rad Mci. sensitivities.
      CR 97-4180 SJAE rad monitor setpoints, addressed coordination problems          ',
      between the Operations procedure used to request new radiation monitor
      setpoints, the Environmental and Radiological Control (E&RC) proced ce
      that calculates the new setpoint, and the Maintenance procedure that
        installs the new setpoints. By the time the radiation monitor setpoints
      were ready to be installed the new values needed to be recalculated.
      The inspector determined as a result of the cited failure and the three
      additional CRs mentioned previously, that control and monitor'.ng of the
      alarm setpoint was poor. Previous instances of failing to properly
      disposition abnormal values were recorded by the NRC in Inspection
      Re) ort (IR) 50-325(324)/97-12, when inadequate corrective action was
      tacen for abnormally high drywell temperature. Tne abnormal temperature
        resulted in exceeding the calculated environmental limits for ten
        snubbers in the drywell.
                                              ~
 
                                  _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ -
    .
  .
                                                                                    13
        c. Conclusions
            The licensee continues to struggle with proper dispositioning of
            abnormal indications. The failure to maintain the Daily Surveillance
            Report in accordance with procedure was a violation. Abnormal values
            observed for the Steam Jet Air Ejector radiation monitor and subsequent
            test indicate potential fuel failure for Unit 1.
      M8    Miscellaneous Maintenance Issues (92902)
      M8.1  (Closed) Licensee Event Reoort (LER) 50-325(324)/96-017-00:                                  Invalid
            Loss of Coolant Accident Locic Actuation
            The invalid LOCA. initiation signal occurred during installation of test
            equipment to support surveillance testing.                                  P16nt systems responded as
            designed. The initiation signal resulted in the following actuation:
                  Automatic start of emergency DGs 1.2.3. and 4.
                  Automatic start of Unit 1 Core Spray (CS) pump 1A.
                  Automatic start of Unit 2 Nuclear Service Water (NSW) pump 2A.
                  Unit 1 Grou) 10 division 1 actuation.
                  Closure of Jnit 1 Reactor Building Closed Cooling Water heat
                  exchanger Service Water isolation valve.1-SW-V106.
                  0)ening of NSW header to vital header isolation valve. 1-SW-V117.
,                  Slutdown of 1A and 10 Unit 1 drywell coolers                                                      ;
1
            Corrective actions, described in the LER. were reviewed and verified by
            the inspector. -These included: appropriate administrative action with
            the involved technician; briefing of maintenance 1&C technicians on this
            event; providing maintenance I&C personnel managements expectations ft
            the restart of surveillance tests after problems have been encountered;
            restricting the use of Simpson Model 260 Voltage Ohm Meters (V0Ms) for
            circuit checks specified in maintenance surveillance tests: developing
            training to enhance technician knowledge of the effects of test
            equipment misalignment: and revising maintenance procedures to preclude
            similar events.
            This event did not violate TS. This LER is closed.
      M8.2 (Closed) LER 50-325/97-009-00: Missed Increased Frecuency Inservice
            Testino Recuirement
            The American Society of Mechanical Engineers (ASME) Boiler and Pressure
            Vessel Code. Section XI, 1980 Edition through Winter 1981. Addenda
            Section IWV-3414(a), requires an increase in test frequency in the event
            an increase in stroke time of 25 percent or more from the previous test
            is observed. Contrary to this requirement, the test frequency was not
            increased as required. The required testing was missed by about two
            weeks. Upon discovery. the valve was tested and the stroke time was
            within the previous value and the test met the ASME Section XI
            requirements.
                                                                                                                        !
                                                                                                                        l
                                                                                                                  ___J
 
                                      . _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _
                                              '
    ,
  ,
                                                                      i
      s
                                                                              14
              The corrective actions to prevent recurrence of this event. described in
              the LER. were reviewed and verified by the inspector. Administrative
              controls have been revised to ensure completed test results are reviewed
            -in a timely manner and changes in test frequency are promptly initiated.
              This event did not violate TSs. This event had minimal safety
              significance from a-valve operability viewpoint since the retest of the
              valve showed it was operable, ASME Section XI provides an intermediate
              condition that allows continued operation without need for immediate
              corrective action. From an administrative view, trending valve stroke
              times is an imaortant indication of valve performance. Corrective
              action taken s1ould improve this situation. This LER is closed.
        M8.3 FClosed) LER 50-325/97-001-00:                                  Rod Block Monitor Surveillance
              .
                  nadeauacy
'
              A discovery that the surveillance procedure fer testing the rod block
              monitor (RBM). did not contain the pro 3er s                                  4
                                                                                              Ncessary to ensure      ;
              testing of the RBM instrument channel 3 int                                '
                                                                                              : tion, This condition
              has existed since November 1996 for Unit 1, ma December 1996 for
'
              Unit 2. Upon discovery, the correct tests were performed on both units
              which indicated that the equipment was in calibration and capable of
              performing its safety function.
              The error was attributed to an inadequate administrative review of                                      ;
              reformatting changes made in September 1996. The surveillance procedure
              changes were being upgraded in accordance with the generic procedure
              writers guide. However, these changes did not insert the proper steps
              to test the RBM inop instrument channel B.
            ' Corrective actions, described in the LER. were reviewed and verified by
            -the inspector. The inspector determined that this event did not violate
              TS since only the test for channel B was missed. The situation was
              corrected within the allowable time specified by TS 3/4.3.4.
              The-results of the RBM inop functional tests performed on toth units
              upon discovery, indicated that the equipment was in calibration and
              capable of performing its intended safety function. This LER is closed.
        M8.4 (Closed) LER 50-325(324)/95-022-00:                                    HPCI System Discharae Flow Element
              Gasket Leak
                During performance of a post maintenance test on the HPCI system. the
              discharge flow element flanged gasket developed a 5 to 10 gallons per
                minute (gpm) leak. Several other problems were also observed with
                system operation.
                Investigation revealed that undersized flange studs had been originally
                installed on the flow element flange, allowing the Flexitallic gasket to
                be installed off center. The off centered gasket degraded during the
                post maintenance test. This condition existed on both units and
                prompted declaring a potential failure of the HPCI system to ]erform its
                intended safety function.                                With the HPCI system inoperable tie TS
                                                                                                                      U
 
                              _ _ _ _ _ _ _ _ _ _ _ .
    .
  .
                                                          15
            oermitted continued reactor operation provide 1 the ADS. CS system, and
            RCIC were operable. This event was withir, Me TS requirement.
            Corrective measures as described in the LER were reviewed and verified
            by the inspector. This LER is closed.
      M8.5 (Closed) Ins)ection Follow-un item (IFI) 50-325/97-05-02:                  Abnormal CS
            Soarcer Brea t Detector Indication
            (Closed) VIC 50-325/97-06-03:                Inadeauate CS Surveillance Procedure
            .(Closed) LER 50-325/97 02:                Core Soray Header Differential Pressure
            Instrumentation InoDerable
            On March 9.1997, en auxiliary o)erator (AO) was verifying
            instrumentation indications in tie Unit 1 Reactor Building.                The A0
            observed.that the reading displayed for 1-E21-PDS-N004A. Core Spray Line
            Break Indicator, was not within TS 4.5.3.1.2.c.2 requirements. This
              )ressure switch functioned to detect a break in the CS piping located
l            3etween the vessel and the shroud. The differential pressure (dP)
            sensor measures the pressure across the core. Due to the addition of
'
L            the drop from the steam separator, any break in the line would cause the
l    -
            indicated pressure drop to increase which would cause a more positive
            indicated dP. The out of tolerance condition had existed since
            November 1996 as stated in LER 50-325/97-02. During review of the
            associated surveillance procedures, the inspector determined that actual
            verification of the CS sparger alarm setpoint in relation to the
            " normal" indicated instrument pressure was not being performed.
L
I
            Themfore. the licensee could not evaluate whether the alarm setpoint
            was within the " normal" TS range. This nonconformance resulted in VIO
            50 325/97-05-02. Inadequate CS Surveillance Procedure.
            The licensee performed reviews of data collected nonroutinely during
              1995-1996 and in ESR 97-181 calculated a " normal" value for setpoint
            verification in the related surveillance procedures. The licensee
            subsecuently changed the alarm setpoints and updated the affected
            procec ures. Additionally. the licensee performed a review of the TS and
            determined that appropriate logging of required TS values was being
            accomplished. During the refueling outage for Unit 2 from Se]tember to
            October 1997 the licensee, with prior NRC approval, uprated t1e 100
            percent _ rated thermal power 5 percent. The licensee included
              verification of CS sparger dP " normal" values as part of the uprate
              test program performed in accordance with S)ecial Procedure 2SP-97-204.
              Unit 2 Power Jprate Data Collection. The cleck served to record the CS
              sparger shutdown values.
              The inspector reviewed ESR 97-634. ESP-97-204. CR 97-3870. LER 50-
              325/97-02, and other related documentation. The inspector verified that
              routine recording. upon entering mode 1. of the CS sparger dP was
              incorporated into 0)erating Instruction 001-03.3. Auxiliary Operator
              Daily Surveillance Report for both units. CR 97-3870. Core Spray Leak
              Detection, documented the discovery on October 29, 1997 by an AD, that
                                                                                                  4
                                                                                                  l
 
                            _ - _ _ _ _ _  - - _ - _ _ _ _ _ _ _
'
                                                                  16
        the 2-E21-PDS-N004A. CS A Loop Leak Detection, was outside of its
        specified range. The instrument was declared inoperable and an LC0 was
        entered. The licensee determined the new CS dP range in ESR 97-634
        Core Spr 3y Loop Line Breuk Detectio , Allowable Range Change. The new
        alarm setpoints were implemented and integrated into the affected
        surveillances. 3rocedures, and design documents. Based on completion of
        the review of t1e TS for other " normal" values not properly trended,
        adjustment of the dP alarm setpoints*to bring the setpoints into
        rvpliance with TS. and the institution of routine monitcring of the CS
        .qarger " normal" values these items are closed.
  M8.6 (Clos (d) VIO 50-325(324)/97-02-04: Failure to Imolement the
        Renuirements of (a)(1) and (a)(2) of 10 CFR 50.65. The Maintenance Rule
        This violation reported that all historical data since July 10. 1993.
        had not been obtained to establish baseline system / structure / component
        (SSC) performance, validate scoping, and set initial condition (a)(1)
        and condition (a)(2) in the case of the reactor protection system (RPS),
        Only corrective work. requests / job orders had been used for initial
        determination of functional failures. Therefore, instrument out-of-
        calibration data had not been reviewed for the period of July 10. 1993
        through October 30. 1995. As an action related to Maintenance Rule
        implementation. Procedure OMMM-004. PM. was revised on October 30. 1995,
        to require that out-of-calibration data be evaluated for Maintenance
        Rule functional failure applicability. However, this requirement only
        collected subsequent instrument out-of-calibration data.
        As corrective action for this violation, the licensee reviewed all
        available instrument out-of-calibration data for the RPS and other
        components / systems which support the Maintenance Rule functions.
        Functional failures identified were evaluated against performance
        criteria to determine whether (a)(1) status should be assigned.
        Although six condition reports were issued to evaluate additional
        functional failures, no system was required to be classified (a)(1)
        based on this review. The inspector reviewed the licensee's corrective
        actions and held discussions with a)plicable management and engineering
        personnel concerning this issue. T1e inspector concluded that the
        licensee had taken the necessary corrective action to correct the
        deficient condition and had taken appropriate corrective action to
        prevent its recurrence.          This item is closed.
                                          III. Enaineerina
  El    Conduct of Engineering
  El.1 Review of Enaineerina Procedures
    a.  Insoection Scoce (37550)
        The inspectors reviewed the licensee's procedures which control the
          environmental qualification program.
 
                            . _ _ _ _ _ _ _ _ _ _ _ _ -
    4
  4
                                                        17
      b.  Observations and Findinas
        :The inspectors reviewed the procedures listed below which control
          various activities related to the environmental qualification 3rogram to
          determine if the procedures implement the requirements of 10 C:R 50.
          Appendix B. and 10 CFR 50.49. The following procedures were reviewed:
          EGR-NGGC 0005. Engineering Service Requests. Rev    6. dated
          Septembe" 5. 1997
          EGR-NGGC-0007. Maintenance of Design Documents, Rev. 2. dated
          August 22, 1997
'
          EGR-NGGC 0153. Engineering Instrument Setpoints. Rev. 3. dated
:        August 22. 1997
l        EGR-NGGC-0156. Environmental Qualification of mlectrical Equipment
l        Important to Safety. Rev. 4. dated October 8.1997
          ENP-13.6 Equipment Data Base System. Control and Revision
          Rev. 12. dated June 25. 1997
          MCP-NGGC-401.. Material Acquisition (Procurement Receiving, and
          Shipping). Rev. 3. dated August 26, 1997
          The inspectors verified that the procedures provided adequate
          instructions for establishing, maintaining and implementing the
          requirements of'10 CFR 00.49 except for the issues discussed
          below.
          Section 9.6 of procedure EGR-NGGC-0156 provided the guidance for
          maintaining E0 qualification data packages (ODPs). The procedure
          specified that changes to ODPs are to be captured using the ESR
          process. The procedure required that ODPs were to be periodically
          updated as necessary to maintain auditability, to incorporate new
'
          requirements, to meet plant specific requirements, ard to keep the
          number of outstanding-changes at a reasonable level. However
5
          procedure EGR-NGGC-0156 did not specify a clear time requirement
          for updating the CDPs. The inspectors also determined that
          procedure EGR-NGGC-0007 did not provide any requirements for
          updating ODPs.    The failure to s]ecify specific criteria in
          procedures could result in the 0)Ps becoming unauditable which is
          contrary to the requirements of 10 CFR 50.49. The failure to
          maintain and u]date the ODPs was one of the causes of the
          violation whic1 resulted in the civil penalty identified in NRC
          Inspection Report (IR) 50-325(324)/96-14. The failure to
          establish clear, definite requirements for updating ODPs was
          identified as a violation example at the Shearon Harris Nuclear
          Plant in NRC IR 50-400/97-12. Since all Brunswick 00Ps are being
          revised and updated at the current time, a violation was not
          identified for this issue during the current inspection. The
          licensee's corrective actions for the Harris plant will resolve
                                                          - ,                    -
 
                            . _ _ - . _ - _ _ - _ - _ - _
    .
  .
                                                  18
      this problem since the Harris. Brunswick, n.d H. B. Robinson
      plants use the same corporate EGR-NGGC ?,ocedures.
      Procedure EGR NG D 0153 provides the methodology to establish
      instrument setpoint margins sufficient to account for various
      instrument uncertainties and environmental effects including
      temperature, pressure, radiation, seismic, and insulation
      resistance errors
      Although procedure EGR-NGGC-0153 provided guidance on the
      treatment of environmental effects, the inspectors noted that in
      the discussion of temperature effects, the applicability of vendor  3
      worst case performance specifications to plant specific conditions  i
      was not clear. The inspectors also noted that requirements for
      seismic effects in procedure EGR-NGGC-0153 were not clear
      regarding t6 match / confirmation of vendor profiles to plant
      specific [      les or configuration,
      in addition, the inspectors noted that procedure EGR-NGGC-0153
      referenced Drawing 0-03056. Service Environment Chart Normal &
      Accident Conditions. Units 1 & 2. for information on accident
      temperature data to be used in instrument setpoint calculations.
      The inspectors determined that-Drawing D-03056 was " frozen" on
      December 12. 1996, and was not available for use. The reason for
      removal of Drawing 0-03056 from use was documented in CR 96-04002
      which identif9d the need to revise. and update Drawing D-03056-to
      incorporate f icironmental data from the Reactor Building
      Environmentai Renort (RBER), Revision 5. The inspectors noted in
      review of calculations initiated since December 1996, the RBER
      was referenced for temperature profiles in the re:ctor building.
      The licensee indicated that a revision to EGR-NGGC-0153 will be
      initiated to resolve inconsistency in wording regarding the
      application of accident temperature / seismic effects to make it
      clear that vendor test results would fully envelope site specific
      profiles unless an evaluation has been aerformed to evaluate the-
      differences. Additional guidance will 3e included to characterize
      the requirements for engineering reviews of test-data to ensure
      seismic and environmental profiles are bounding for site specific
      conditions. The licensee indicated procedure EGR-NGGC-0153 will
      also be revised to either remove D-03056 as the reference for
      temperature data and replace it with the appropriate reference
      (the RBER)  . or to correct the drawing.
      The inspectors also identified that procedure EGR-NGGC-0153 unde-
      Section 9.5.1. Calibration Errors, was not clear regarding
      instrument calibration surveillance requirements for as-left, as-
      found or leave-alone zone tolerances. The licensee indicated
      that procedure EGR-NGGC-0153. Section 9.5.1. would be revised to
      clarify these requirements to indicate that calibration tolerances
      are the defined limits, above and below a desired value, within
      which an instrument loop signal may vary and not require
a
                                                                            j
 
                      _..__-____- ___-_____-__ _____- -
                                                                                                    .
                                                                                                            ..
                                                                                                                                ..
  4
.
                                                                      19
          adjustment.                                Licensee engineers stated that calibration tolerances
          are understood to be "as-left" values.
          The inspectors will review Procedure EGR-NGGC-0153 in a future
          inspection to followup on these issues. An ins)ector followup
          item (IFI). 50-325(324)/97-13 06. Revisions to )rocedure EGR-NGGC-
          0153, was identified to the licensee pending further review by
          liRC.
      c.  Conclusions
          With the exception of the issues discussed above, the inspectors
          concluded that the licensee's procedures for implementation of the
          Environmental Qualification com) lied with the requirements of 10 CFR
          50.49 and 10 CFR 50. Appendix 3. An IFI was identified to review
          procedure EGR-NGGC-0153 to verify that the licensee incorporates the
          above comments and clarifications. The reference to a " frozen" drawing
          to obtain accident temperature data and the wording inconsistencies
          discussed above were identificd to the licensee as a weakness.
    El.2 Review of Instrument Setooiit Calculations
      a.  Insoection Stone (37550)                            ,
          The inspectors reviewed randomly selected instrument setpoint
          calculations to deternine the adequacy of the licensee's calculations.
      b.  Observations and Finninos
          The inspectors reviewed the instrument setpoint calculations
          listed below and verified that the calculations were completed in
          accordance with NRC requirements. The inspectors verified that
          the calculations incorporated industry standards. Updated Final
          Safety Analysis Report commitments. Technical S)ecification
          requirements, and recommendations contained in iRC Regulatory
          Guides.    Calculations reviewed were as follows:
          -
                  -Calculation OE41-0036. Power Uprate HPCI Steamline Flow High
                  Uncertainty and Scaling Calculation.
            -
                  Calculation ORWCU-0010. U1/U2 RWCU Flow Accuracy
                  Calculation. Units 1 and 2 RWCU Differential Flow Leak
                  Detection / BESS I&C.
            -
                  Calculation 0821-0068. Power Uprate Main Steam Line Flow
                  High Setpoint Uncertainty and Scaling Calculation.
            -
                  Calculation 0-01534A-297. Insulation Resistance Degradation
                  Calculation.
          From review of System Description SD-01.2. Reactor Vessel
            Instrumentation. and the Safety Evaluation by the Office of
                                                          ,                                                  _ _ _ _ _ _ _ _ _    a
 
                  _ _ _ _ _ _ _
                                      20
  Nuclear Reactor Regulation. Conformance to Regulatory Guide 1.97
  Revision 2. Brunswick Steam Electric Plant. Units 1 and 2. Dated
  May 14. 1985. the inspectors concluded that these calculations
  were. typical. The instrument setpoint calculations typically
  considered 140 F as the maximum temperature in the calculations.
  From review of the calculations, the inspectors determined that
  instruments that perform a safety function are analyzed for a LOCA
  environment in the reactor building. The calculations showed that
  instrument uncertainties considered instrument temperature effects
  for a maximum temperature of 140' F which is bounding for the
  analyzed LOCA environment.
  The inspectors also determined that instruments relied upon to
  mitigate the effects of a high energy line break (HELB) were also
  evaluated by the licensee. For this instrumentation,
  environmental uncertainties-for a harsh environment were not
  required to be considered since the instrumentation function would
  occur before the reactor building temperature )rofiles listed in
p  the Reactor Building Environmental Report (REBR) Revision 6.
  dated November 5. 1997, would reach 140 F and affect instrument
  performance. The ins)ectors noted that abnormal temperatures were
  not discussed in the-RBER. Discussions with licensee engineers
  disclosed that the design base accident event is based on an
  initial building environment airspace temperature of 104 F. The
  building temperatures ace measured and recorded daily by plant
  operators in accordance with procedure numbers 101-03.4.1 and 201-
  03.4,4. Unit 1 and 2 Control Operator Daily Check Sheets.. The
  = operators are required to contact the duty engineer when the
  reactor building temperature exceeds 104 F so that engineering
  can perform an assessment of the effects of temperature on
  environmental qualification.
  -The inspectors noted that calculations for instrumentation which
  mitigates a HELB demonstrated that the instrument and associated
  equipment would not be exposed to a harsh environment before the
    instrumentation performed its safety function. In the instrument
  calculations reviewed by the inspectors instrument setpoints were
  based on a maximum temperature of 140 F (non-steam environment).
  Although allowances were not made for a harsh environment. a
    seismic allowance was included in the calculations.
    Review of the temperature profiles as shown in the Brunswick
    Reactor Building Environmental Report showed that the actuation
    isolation signal would occur before exceeding the temperature
    allowances assumed in the setpoint uncertainty calculations. An
    exce) tion was the High Pressure Coolant Injection (HPCI) line
    breat in the steam tunnel where the temperature profile showed
    that 140 F would be exceeded for ap3roximately 2.5 seconds before
    the isolation trip _ signal occurs. iowever this instrumentation
    would remain operable based on thermal delays. However, the HPCI
    isolation function would most likely be initiated by temperature
 
    .
                                                      21
                  sensors in the steam tunnel or HPCI room which would occur
                  imediately with no time delay.
                  The inspectors concluded that the instrument setpoint calculations
                  complied with NRC requirements and were technically adequate.
                  Review of the calculations showed that environmental effects,
j-                specifically accident temperature, were correctly evaluated in the
                  calculations,
            c.    Conclusions
                  The inspectors concluded that the licensee's calculations were
                  technically adequate and complied with NRC requirements.    The
                  inspectors concurred with the licensee's conclusions that the
                  setpoints for instruments relied upon to mitigate the effects of a
                  KLB did not require inclusion of uncertainties for a harsh
                  environment since the instruments perform their ft..iction before
                  being effected by the harsh environment. Setpoints for
                  instruments required for LOCA effects include the appropriate
                  environmental uncertainties.
        -El.3 Enaineerina Service Reaucst (ESR) 97-00426
            a.    Inspection Scoce (375501                                                    '
                  The inspectors reviewed ESR 97-00426 which was prepared to address
                  questions on instrument setpoints.
            b.    Observations and Findinas
                  A review of procedures and various documents by an independent
                  consultant resulted in questions involving environmental effects
                  including uncertainties on instrument accuracy. These guestions          .
                  were dccumented in an E-mail message dated June 20, 1997 Subject:        '
                  E0 and Instrument Accuracy. The licensee addressed the referenced        !
                  memo in Engineering Service Request ESR 97-00426. Revision 0.
                -dated September 18. 1997. ESR 97-00426 documents the evaluation
                  completed by the licensee to address environmental effects on-
                  instrumentation. The inspectors noted that the licensee response
                  did not address the questions in the June 20, 1997 E-mail message
                  point by point. but provided an evaluation that was more generic
                  in nature. The inspectors noted that ESR 97-00426 was an
                  engineering disposition (ED) type ESR. as defined in procedure
                  EGR-NGGC-0005.  The use of this type ESR to respond to the E-mail
                  cuestions was appropriate since the ESR only communicated existing
                  cesign requirements, did not produce design output, and did not
                  change existing engineering documents.
                  The ESR concluded that instruments that aerform a safety function
                  are analyzed for a LOCA environment in t1e reactor building. The
                  instrument uncertainties consider -instrument temperature ef fects
                    for a maximum temperature of 140"F which is the maximum bounding
  . . .      .
                    ..
                                              . . . .    .
                                                                .
                                                                          .
                                                                                    .. - .
                                                                                          o
 
                          _ _ _ - _ - .    _ _ _ _ _ _ _ .          .
                                                                          .
                                                                                _        ..
    .
                                                              22
            temperature for the analyzed LOCA environment.                The inspectors
            noted that the word minimum had been incorrectly used in the
            fourth line, third paragraph in Section 2.0 of the ESR. The
            licensee stated that they will correct this error when the ESR is
            revised. as discussed below.
            ESR 97-00426 also concluded that harsh environmental effects have
            been appropriately accounted for in safety related uncertainty
            calculations. The ESR concluded that the isolaticr. aquence for a
            HELB due to main steam line break. reactor core isolation cooling
l            steam-line break, high pressure coolant injection steam line
            break, cr a piping failure in the reactor water cleanup system is
            such thtt the isolation function will occur before the
            instrumentation is exposed to harsh environmental effects. This
            conclusion was based on the instrumentation being able to perform
            its safety function prior to the temperature exceeding the
            temperature allowance assumed in the setpoint calculations. For
            area temperatures exceeding the setpoint temperature uncertainty
            allowance, the use of emergency operating procedures (EOPs),
            operator action, and local temperature instrumentation would
            mitigate the event and provide the actions to determine and/or
            maintain. reactor level during a LOCA or HELB.
            When temperatures exceed the temperatures (140 F) assumed in the
            setpoint calculations, plant operation is controlled through the
          ' COPS.    A review'of E0P-03-SCCP Revision 5. Secondary Containment
            Control Procedure, and 2EOP-LPC Revision 1. Level / Power Control,
            shows that high area temperatures are an entry condition into
            secondary containment control procedure E0P when area temperatures
            exceed the maximum safe operating value requiring manual reactor
            sCrdm.
            E0P-03-SCCP Revision 5. refers the operators to Caution 1 to
            determine reactor level instrumentation operability. A review of
            Caution 1 disclosed that vessel level wide range instrumentation                        ;
            8B21 - LI - R604A/604B and C32 - PR - R609 are not to be used when
              secondary containment temperature exceeds 140 F. This exclusion
            was because the reference leg and associated instrumentation for
              these loops are in secondary containment. E0P Caution 1 then
              )rovided compensation data for the remaining level instrumentation
              ]ased on drywell tem]erature, reactor saturation limit, and
              reactor pressure.          iowever, for secondary containment
              temperatures above 140 F. Caution 1 instrumentation may not be
              o)erable with instrumentation exposed to temperatures greater
              tlan 140*F during an event. In cases when vessel level can not
              adequately be determined, the E0Ps direct the operators to
              depressurize by initiating ADS and flood the vessel using low
              pressure emergency core cooling systems.
  .
      . ..
                      .. . .
                                                            .. .
                                                                  . .
                                                                        .
                                                                                        .    .
                                                                                                . . .  ,
 
              - _ _ _ _ _ _ _ _ _ _ _ _ - _                      _.
                                                                                  .
                                                                                                            ..      .
      ..
                                                                        23
              c.                    Conclusions
                                    The inspectors concluded that the licensee adequately addressed
                                    the questions in the June 20. 1997 E-mail message regarding
                                    instrument and E0 accuracy. However, the licensee stated that
    _
                                    they will revise F.SR 97 00426 to address each question and
                                    recommendation ir. the E-mail message point by point to further
                                    clarify their response to the concerns / issues raised in the
                                    June 20, 1997 E mail message.
            El.4 Environmental Qualificat%1
,
              a.                      Insnection Scooe (37550.92903)
                                    The inspectors reviewed the licensee's corrective actions for the
:
                                    Environmental Qualification (FO) program, in response to findings
l                                    identified during Self-Assessment numbers 95-0041 and 96-0271 and
                                    the violations identified in NRC IR 50-325(324)/96-14.
              b.                    Observations and Findinas
                                      1) Review of E0 Equipment Data Base
                                    The licensee's corrective actions to resolve the discrepancies in
                                    the E0 program identified by NRC (See IR 50-325 324/96-14)
                                      include corrections to and updating of the Equipment Data Base
                                    System (EDBS). Numerous errors in EDBS had been identified and
                                    corrected by the licensee since the inspection findings were
                                      identified in IR 50-325(324)/96-14. The errors in EDBS were                      .
                                      identified during E0 equipment walkdowns and review of various                  !
                                      data bases. In addition, numerous errors were identified in the
                                      EQ zones listed in EDBS for the location where various components
                                    were installed. These primarily occurred at. zone boundaries and
                                      were being resolved during review of walkdown data.
                                      The requirements for. recording and correcting E0 data in EDBS was
                                      s)ecified in- CP&L procedures EGR-NGGC-0156 and ENP-33,6.  The
                                  -c1anges to EDBS to correct errors were processed using Form 100 of
                                      ENP-33.6. The Form 100 was design verified in the E0 unit and was
                                      then forwarded to appropriate personnel for entry into EDBS. All
                                      EDBS data entries made were independently verified by personnel in
                                      the Configuration Management group in the Design Control Unit.
                                      The independent verification was performed to minimize o-
                                      eliminate data entry errors. Additional corrections to EDBS were
                                      ongoing to incorporate E0 walkdown ins)ection results and the
                                      revisions to EQ qualification data paccages.
                                      The inspectors reviewed some randomly selected revisions to EDBS
                                      identified as a result of the E0 corrective actions and verified
                                      the EDBS data had been corrected. The inspectors also discussed
                                      the program for control of changes to EDBS with various licensee
                                      personnel who perform the day to day system revisions. These
  .
        ..                    .
                                            .. .    .
                                                                .    .
                                                                                                          ..
                                                                                                                ____
 
              _ _________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
                                                              24
  discussions disclosed that these individuals were cognizant of the
  requirements for controlling and making corrections to EDBS.
  2) Review of Qualification Data Packages
  The inspectors reviewed a draft cop.f of Revision 4 of ODP No. 49.
  titled. " Qualification Data Package For NAMCO EA180 Series Limit
  Switches" to determine if it adequately demonstrated environmental
  qualification for the safety related NAMCO switches for use inside
  the drywell in accordance with 10 CFR 50.49 and appropriate
  licensee E0 Prccedures. The package addressed the following:
  qualification level (0588 Cat. I); tag numbers of equipment
  covered in the QDP: test report aaplicability; similarity of                          test
  specimens to installed equipment: E0 parameters. temperature,
  pressure, relative humidity, radiation, chemical spray,
  submergence; cualified life: E0 maintenance requirements; test
  anomalies; anc operating experience items.
  During review 3f the Draft ODP. the inspectors identified the following
  questions / comments:
  .
          The text in the CDP indicates that there were five anomalies in
.
'
          Qualification Test Report (OTR) 130 but only four anomalies were
          discussed in the ODP.
l .
          Attachment 2 to the ODP included a calculation for qualified life
l        of the limit switches which was not signed as reviewed.
  *        Differences were noted in the system component evaluation
          worksheets (SCEW) for the same limit switches in the different
          units.
  *        Data was missing from some of the SCEW sheets. That is, there
          were blanks on the data sheets.                        For example, data on accuracy was
          left blank.
  *
          Some components were specified with Anaconda flex and others just
          stainless steel flex conduit. Additionally, only certain
          components were specified for weep holes.
  *      Page 49 section 4.1 Installation requirements indicates that the
          conduit seal may not be necessary for those limit switches
          installed in the Reactor Building. This requirement should be
          clear and should specifically list those limit switches which
          require conduit selling to ensure qualification.
  *      Page 13 lists the 16 Namco EA180 limit switches which had been
          installed. However only 14 were considered qualifieo by this ODP.
          Unit I limit switch tag numbers 1821-ZS-5373 and 1B21-ZS-5374 were
          excluded from the E0 requirements by ESR-97-00431. The Unit 2
          equivalent switches were not discussed in the ODP.
                                                            .                                      _ _ _ _ .
 
                    . _ _ _ _ _ _  _ _ _ - _
                                  ,
                                              25
*      In Section 2 of the 00P it was stated that it was a good
        maintenance practice to lubricate the NAMCO limit switches.
        however. lubrication was not specified in Section 4 of the ODP
        which lists recommended maintenance practices.
*
        In Section 4.2 of the ODP it was stated that the switches can be
        refurbished. However, a statement was made on page 21 that
        qualified replacement part kits were no longer available.
*      A reference was made to abnormal temperatures on page 38 of the
        ODP.  However, abnormal temperatures were not included in DR 227.
*      The inspectors questioned apparent inconsistencies between
        activation energies and aging methods discussed in referenced
        qualification test reports (OTRs).
The licensee indicated that these comments would be evaluated by
the E0 group and if appropriate, addressed in Revision 4 of the
QDP when it is completed.
The inspectors reviewed a draft copy of Revision 7 of ODP-67
General Electric Company IC 7700 Series Motor Control Centers for
BNP.    The GE MCCs. located or, the 20, 50, and 80 foot elevations
of the Units 1 and 2 Reactor Buildings, are subject to harsh
environments resulting from postulated design basis accidents and          ;
have a safety function to mitigate the consequences of these              F
accidents.    The MCCs were qualified in ODP-67.
A series of similarity analysis were performed to demonstrate
similarity between the tested configuration and supplied. The
inspectors reviewed portions of DR 232. "Nutherm Report No. CPL-
7806R. Qualification Test Results Applicable to Brunswick Nuclear
Power Plant Safety-Related GE 7700 MCCs." Revision 0, dated
June 30, 1997 which dccumented the similarity analysis. Section 2 of
DR-232 contains a discussion on the similarity analysis between the
components tested by NUTHERM and those installed in the Brunswick MCCs.
The similarity discussion covers fuses, stab assemblies, control
transformers control and power wiring, overload heaters. overload
relays, terminal boards, starters and contactors, molded case circuit
breakers. circuit protectors. disconnect switches. potentiometers, and
indicating lights. The similarity analyses were based on the similarity
analyses contained in DR 1.1. GE Company NEDC-30696-P. May 1985. MCC
Oualification Test Report Phase Il for CP&L Brunswick Plant, or were
devices which could be directly linked to a test specimen and did not
require a similarity analysis. Based on review of DR-232 NRC concluded
that NOTHERM was able to establish that the com3onents they tested were
  in the same family as those provided by GE in t1e MCCs. This review was
also dccumented in IR 50-325(324)/97-09.
A draft copy of Revision 0 of ODP 99. R. G. Laurence Series 500
and 600 Solenoid Valves was reviewed. The inspectors verified
that similarity analysis was included in the ODPs.
                                                                          >
                          +
 
                _ ___ __        ___ - __ _        _        -.
  .
                                            26
    3) Review of EO Walkdown Data
    The inspectors reviewed E0 walkdown data which document inspection
    of E0 equipment-in the Unit 2 MSIV pit and drywell, and Unit 2
    reactor building. Tha E0 walkdowns were performed in accordance
    with CP&L Special Procedure OSP-96-014. EQ Equipment Field
    Verification. The pyrpose of the walkdowns was to verify the
    accuracy of the manulacturer/model number listed in the licensee's
    data bases and to verify the equipment installed orientation and
    configuration were in accordance with the E0 qualification
    documentation.        The ins)ectors reviewed walkdown records for scram                      '
    pilot' solenoid valves, 1AMC0 limit switches, temperature elements,
    excess flow check valves, and pressure switches. The walkdown
    data was recorded on field inspection data sheets which were'then
    converted into an electronic data base. The inspectors verified
    that discrepancies identified during the walkdowns were documented
    either on a work request (WR/J0) for repair, or in a condition
    re) ort (CR).        The ins)ectors reviewed completed WR/JO numbers 97-
    AF  JR1, 97-AFUR2, 97- A UR3, and 97-AFUR4. These WR/J0s document
    drilling of weepholes in junction boxes in the Unit 2 MSIV pit to
    resolve a moisture intrusion issue. These boxes are associated
.
    with limit switches for the Unit 2 main steam isolation valves.
L  The completed WR/J0s showed that the weepholes were drilled to
    resolve the concerns. The inspectors did not identify any
;  discrepancies in the records reviewed.
    4) Review of Environmental Qualification Condition Reports
    The inspectors reviewed the licensee's corrective c.,ctions to
L  disposition the CRs listed below. These CRs were initiated by the
    licensee to-document and disposition nonconforming items whicn
    were identified during the ongoing E0 reconstitution project. The
    nonconforming items were identified as a result of E0 equipment
    walk h ns, review cnd updating of E0 equipment ODPs, omissions
    from the original program, or changes to the operating
    environment. The CRs reviewed were as follows:
    CR 97-02015
    The licensee initiated CR 97-02015 on June 6. 1997 to document and
    disposition deficiencies that had been identified by the
    licensee's training staff during observation of simulator training
    when the fire protection system had not been isolated within the
    15 minute time period after initiation of a HELB specified in
    31 ant o)erating 3rocedures. The 15 minute time period is the
    ) asis w1ich esta)lished flood '.evels for E0 e
    and north and south RHR and core spray rooms.quipment              in the HPCI
                                                            Review of closure
    for CR 97-02015 disclosed that the licensee concluded that the
    issue has been adequately addressed by operator training,
    primarily through critiques which were held following the
    completion of the simulator training to discuss deficiencies noted
    during the training. In response to the CR. Action Items were
Y
                                                                                  _ _ _ _ _ _ _ _
 
  - -. - - -          -          __ - . - - - - .
              .
1
                                                      27
'
                assigned to the Operator Training group to incorporate the basis
,
                for the need to isolate the fire protection system into training
                materials. However, review of the training records on June 12,
                1997, by personnel from the E0 group resulted in additional
                questions regarding the licensee s corrective actions. The
                records reviewed by the E0 personnel indicated that during
'
                simulator training, approximately 10 to 20 percent of the
                operators were failing to enter AOP-05,0, Radioactive Spills. High
.                Radiation, and Airborne Activity, or were entering the AOP late
;                (after 15 minutes). The inspectors made an indepen6nt review of
,                the training records reviewed by the E0 personnel. This review
                disclosed that the records the E0 personnel reviewed on June 12,
                1997 were for the six month
                02015 (January - June 1997) The    .
                                                      period prior to reviewed
                                                          inspectors  initiation of CR 97-
                                                                                training
                records for July - September, 1997 and noted significant
                improvement in this area, although the HELB scenario was not
                included as part of the simulator training exercises in this time
                period. The training scenario did include a torus leak which
                required entry into A0P-05.0.
I
                The inspectors noted that the concern regarding flooding of
_
                  instruments could also be caused by other accidents such as pipe
L                breaks in the service water or Reactor Building Closed Cooling
l                Water (RBCCW). Operator actions in these cases would be directed
e                by E0P-03 SCCP Secondary Containment Control Proccdure (SCCP),
                based on high water leve'is in the HPCI and north and south RHR and
                core spray rooms. An uttry into E0P-03-SCCP would also result
                  from flooding in these same rooms caused by activation of the fire
                protection system. As aaditional followup on this issue, the
                inspectors observed simulator training scenarios performed on
                December 3 and 17, 1997. Included in the scenario was a RCIC
                steam line break (HELB) and activitation of the fire protection
                system. Both crews participating in the training scenario
                isolated the fire protection system within the 15 minute time
                period. The inspectors also questioned some randomly selected
                reactor operators regarding the need for entry into A0P-05.0
                following a HELB. The operators were cognizant of the basis of
                the actions in A0P-05.0 (need and reason for isolating the ' ire
                protection s
                CR 97-02015.ystem) and were familiar with the problem addrc ses by
                The inspectors verified the action items associated with the CR
                were completed.      CR 97-02015 was closed on December 11. 1997.
                CR 97-01841. 97 02025. & 97-02408 These CRs documented various
                issues regarding possible effects of moisture on E0 equipment. CR
                97-0184) was initiated to document the effect of spray from the
                fire protection system on E0 equipment in the reactor building.
                The licensee has resolved all the issues associated with this CR
                except for drilling of weepholes in junction boxes whicn may be
                affected by the water spray. Licensee engineers are currently
                - preparing instructions and procedures for completing this work.
 
                                _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
                                                                      -
  .
                                                                    28
    The problem documented in CR 97-02025 concerned an issue which had
    been the subject of IE Circular 79-05. Moisture Leakage in
    Stranded Wire Conductors, which was issued by NRC on March 20.
    1979. This affects Patel seals which were used to seal some
    stranded wire conductors in instrument circuits. CR 97-0?408
    documents several other moisture intrusion issues. The immediate
    corrective action taken to resolve these issues, as documented in
    CR 97-02408 was to hire an outside consultant to address the
    issues. The consultant has reviewed many of the issues documented
    in CR numbers 97-01841, 97-02025. and 97-02408 and made
    recommendations, some of which have been implemen.ed. The
    consultant also addressed another issue in the CRs involving
    current leakage in control circuit and the possible impact on ODPs
    and E0 of equipment. This concern was the effect of moisture
    intrusion through stranded wire conductors, sealed with Patel
    seals, which could result in leakage currents in instrument
    circuits. ESR 97 00440 was issued for the 120 volt AC circuits and
    ESR 97-00441 for DC circuits. These ESRs are currently being
    reviewed by licensee engineers. The current leakage issue was
    also applicable to questions raised regarding the NAMCO limit
    switches. The inspectors will review the licensee's evaluation of
    current leakage and its ap311 cation to evaluation of E0 equipment
    in a future inspection. T11s was identified to the licensee as
    IFl 50 325(324)/97-13-07. Review Technical Evaluation of Current
!
    Leakage and the Effect on EQ Equipment. pending further review by
    NRC.
    The licensee also aerformed an evaluation of the potential for
    moisture wicking t1 rough Patel seals. This evaluation was
i  documented in ESR 97-00423. 03erability Evaluation - Wicking.
    Review of the ESR disclosed t.at the licensee performed a detailed
    evaluation of the Patel seals by comparison of the installations
    at Brunswick with the configurations tested by NRC at Sandia
    Laboratorics (NUREG/CR 0699. Jublished March.1979).                  The
    licensee's conclusions were t1at the design function of the
    instellea equipment will not be effected by moisture intrusion
    through the stranded wire. The ESR was based on a review of the
    duration of the design accidents and the resulting leakage
    currents caused by moisture intrusion into limit switches.
    Further review of this ESR will be performed as part of IFI 50-325
    (324)/97-13-07, discussed above.
    CR 97-02016 & 97-02074
    CR numbers 97 02016 & 97-02074 were initiated to document issues
    involving NAMCO limit switches. The following issues were
    identified in the CRs:
    *      Inability to identify the date of manufacture of the switches
            since the codes for date of manufacture were painted over.
 
                              __ ________ ____
                                                      __ _ _
I
                                              29
  *      Potential for paint to impair the operability of the switches.
        The concern was that paint on the roller arms would impair
        mechanical function of the switches.
'
  *      Chemical reaction between paint and internal switch components
        would cause corrosion of switches, leading to failure of the
        switches.
  *      Use of incorrect qualification test reports (0TRs) in the
        qualification test reports which qualified the switches.
  *      Effect of current leakage on switch operability.
  A total of 14 NAMCO limit switches were covered under the E0
  program. These switches were installed during modifications
  completed in 1983 and 1984 The licensee has determined that none
  of the switches were purchased or manufactured prior to 1980.
  Therefore, the concern raised by IE Bulletin 79-28. Possible
  Malfunction of NAMC0 Model EA 180 Limit Switches at Elevated
  Temperatures, would not apply to the switches installed at
  Brunswick, Review of the licensee's response to IEB 79-28
  disclosed that none of the potentially defective switches had been
  purchased by the Brunswick site.
  Review of the i1censee's corrective actions completed to date
  disclosed that the following actions have been completed:
  *
        The licensee has identified the date of manufacture for most of
        the NAMCO limit switches.              Additional manufacture dates may be
        identified when the Unit 1 walkdowns are completed during the
        Spring 1998 refueling outage. However, the licensee has
        conclusively determined that none of the switches would be
        affected by the defects identified in IEB 79-28.
  .
        The switches were stroked in accordance with frequencies per the
        Technical Specifications which demonstrates that the mechanical
        function of the switches had not been impaired by the paint.
  *      The paint has been tested.              The test results show the
        not cause corrosion or deterioration of the switches paint would
  .      The ODP. has been revised to incorporate the correct OTRs. The
        ODP. ODP 49, was still in draft.
  .      The current leakage issue has been evaluated " ESR numbers 97-
        00440 and 97-00441, which are currently being reviewed by licensee
        engineers.
  The licensee subsequently has determined that the switches were
  still within their qualified ;'fe. No equiament operability
  issues related to tv.e NAMCO ilmit switches lave been identified.
                                                                        _
 
.
                                    30
  [R 97 02367
  This CR was initiated on July 3, 1997 to document the failure to
  initiate CRs for nonconforming items, specifically, MCC door
  gaskets and non standard Raychem splices identified as a
  violation by NRC during an inspection documented in NRC 1R 50-325
  (324)/97 08.- The licensee's corrective actions included
  completion of a review of all the E0 walkdown data sheets to
  identify any nonconforming equipment. Additional corrective
  actions included training of personnel in the E0 group regarding
  the corrective action program and assessment of the effectiveness
  of the corrective actions. These correcthe actions were also
  associated with other similar corrective action CRs. such as CR
  97 01972 and CR 97-02465. The inspectors reviewed the completed
  corrective actions and concurred with closure of CR 97 02367. The
  CR was closed on December 14. 1997.
  CR 97-02465 and 97-02672
  This CR wac initiated on July 15, 1997, to document concerns on EQ
  operability determinations.    This CR referenced CR numbers 97-
  01841, 97 02025. and 97 02408. discussed above, which involve
  moisture intrusion issues. As a result of the concerns raised in
  CR 97 02465, the E0 group presented an action plan to resolve the
  moisture intrusion issues (CR 97 02465) to the plant nuclear
  safety committee. Although, further review showed the operability
  determinations for the three CRs were correct, the root cause
  analysis concluded that there were other problems which resulted
  in CR 97-02465.
  The root cause of CR 97-02465 was attributed-to weak E0 project
  management. The root cause/ event review for the CR listed the
  causal-factors indicative of weak E0 3roject management to be poor
  communications within the E0 group, tie site position that E0
  problems were primarily docunitation problems, and a poor
  corrective action culture within the E0 group. The poor
  corrective action culture was evidenced by corrective action items
  which were routinely extended, overdue, or completed late: failure
  to prepare JCOs: numerous CRs written against the E0 grou) for
  improper corrective actions: and closing CR action items )y other
  action items without completing the corrective actions. A
  violation of NRC requirements was identified in IR 50 325, 324/97-
  12 for failure of the licensee to implement their corrective
  action program.
  The licensee's corrective actions to address the issues raised in
  CR 97-02465 included increased management oversight    aerforming a
  review of the E0 project schedule to complete the higlest priority
  work activities first, conducting more frequent E0 group meetings
  to improve communications within the E0 group, transferring some
  E0 group functions from the Design Control l%1t to a site
  organization. and performance of an effer' ve. dss review of the
 
                                                                                  .
            .
          .
                                                31
              completed corrective actions. The CR was closed on December 17,
              1997. The inspectors reviewed the completed corrective actions
              .and concurred with closure of the CR. The ins)ectors concurred
              with the licensee's conclusions that the opera]ility
              determinations for the three referenced CRs were appropriate. NRC
              will perform review of the liccasee's actions to correct the        l
              violation in future inspections,
              CR 97-02672, which was inniated on August 5. 1997, indicated that
              the Supervisor comments listed in CR 97 02465 were a misstatement
              of the consensus of opinion of individuals which met to discuss CR
              97-02465. Review of CR 97-02672 disclosed that the CR did not
              raise any new issues or conceriis which had not been addressed by
              CR 97 02465. CR 97-02672 was closed on December 17, 1997.      NRC
              concurs with the licensee's conclusions and closure of the CR.
              CR 97 4059
              This CR was initiated on December 2, 1997, to document concerns
              and questions on ESR 97-00426. The questions involved
              appropriateness of E0P actions, the need to include evaluation of
              drywell instrumentation in tic ESR, and various questions on
              instrument setpoints. The 1 Lensee completed a review of the
              questions raised in the CR and concluded that the ESR had
              addressed these issues, or the issues were beyond the scope of the
              ESR,  For exam)le, appropriateness of E0P actions were approved by
              NRC for all BW1s and do not involve instrument setpoints. There
              are no instruments in the drywell which provide signals for
              automatic actuation. The inspectors reviewed the licensee's
              responses to the questions in the CR and concurred with the
              licensee's conclusions that no new corrective actions were
              required to resolve the concerns / questions raised in CR 97-04059-
              which had not been previously resolved.
              5) Review of Environmental Qualification Requirements in
              Procurement Practices
              Th'e inspectors reviewed CP&L procedure MCP-NGGC-0401, Material
              Acquisition (Procurement. Receiving, and Shipping). Revision 4,
              dated August 26, 1997. This procedure specifies the instructions
              for procurement of safety related materials for use in CP&L
              nuclear plant. The inspectors noted that the requirements for
              obtaining reviews by E0 engineers is specified in the procedure.
              Discussions with licensee engineers and review of previous
              revisions of the procedure disclosed that the procedure had been
              revised to strengthen the need for the E0 review in Revision 2 of
              .MCP-NGGC 0401, effective April 15. 1997. Revision 2 added
- - - - -
              requirements that components that require environmental
              qualification:be reviewed by the E0 group.
              During review of CRs. the inspectors identified several examples
              of acceptance of materials / equipment by procurement engineering
 
                  _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
                                                            32
  for use in E0 installations which were based on test reports which  ;
  had not been reviewed by the E0 group -These were documented in
  'CR numbers 97-01970 and 97-03036, Several additional examples of
  discrepancies in documents prepared by procurement engineering
  which affected E0 equipment were also identified during review of
  procurement specifications and other documents su * as material
  evaluations. These discrepancies were documented in CR 97-04035
  which tas initiated on November 25, 1997. The review of
  procurement documents was being performed as part of the
  corrective actions to address the E0 program discrepancies
  identified in IR 50-325(324)/96 14. This was listed as Commitment  *
  #4 in the licensee's December 19, 1996 Reply to Notice of          ,
  Violation,
  6) Equipment Lubrication Requirements
  The inspectors reviewed CP&L procedure MMM-053. Equipment
  Lubrication Application Guidance and Lubricant Listing,
  Revision 6 dated November 11, 1997. This procedure provides a
  listing of plant equipment with recommended lubricants to be used,
  guidelines for lubrication of plant equipment, and lubricant
  sampling methods. The inspectors identified the following issues
  after reviewing the procedure:
  -
          ODPs 26, 68, and 88 were not referenced in procedure MMM-
          053. These ODPs cover environmental qualification of
          Reliance electric motors.
  -      Document References corresponding to above ODPs were not
          referenced.
  -      The types of lubricant specified fo, the Reliance motors in
          procedure MMM 053 differ from those listed in the ODPs 26
          and 68.
  -
          Procedure MMM 053 permits maintenance to change the
          lubricant without obtaining engineering review or approval.
  Discussions with licensee engineers disclosed the CR 97-04015 was
  initiated on November 20, 1997, to document the fact that the
  procedure permits changes to lubricants without performance. of an
  engineering review. Action Item 40 to CR 97-02627 was issued to
  document a similar issue. This action item was closed by CR 97-
  04015.
  The inspectors determined that the licensee had not evaluated that
  the type of lubricants (Mobil) specified-in procedure MMM 053 for
  Reliance electric motors differed from those listed in ODP 26 and
  68, Review of ODP 26. Revision 1. Joy Fan /Peliance Electric
  Company, Class 1E Continuous Duty, 20 HP, and ODP 68, Revision 5.
  -Standby Gas Treatment System - Fair Company Filter Unit and
  Control, showed that the electric motors were both qualification
 
                                _ _ _ _ _ - _ _ _
  .
                                                  33
'
          tested using Chevron SRI 2 grease.        The impact of using a
          dif ferent type of grease to lubricate the motors on the
          environmental qualification testing of the motors had not been
          documented by the licensee. The licensee initiated CR 97 04064 to
          document the fact that substitution of alternate lubricants had
          not been evaluated by E0 engineers. The failure to establish
          maintenance procedures appropriate to the circumstances for
          performing maintenance was identified to the licensee as another
          example of violation item 50 325(324)/97-13-02. Inadequate
          Procedure for the Conduct of E0 Preventive Maintenance.
      c. Conclusions
1
          One violation example was identified regarding an inadequate E0
          maintenance procedure for lubrication of E0 electric motors. Two
          inspector followup items were identified to followu) on revisions
          to instrument setpoint procedures and to review leacage current
          calculations. The licensee was making progress in resolving and
          closing CRt identified by the E0 group. As of the inspection
          dates, no 0DPs had been issued.
    E5    Engineering staff Knowledge and Qualification
    E5.1 Trainino and Qualification of E0 Personnel
      a.  Insnection Scone (37550)
          The inspector reviewed the licensee's program for training and
          qualification of personnel in the E0 task force. including both
          CP&L and contract engineers,
      b. Observations and Findinos
          The requirements for performance of E0 equipment walkdowns are
          specified in CP&L Special Procedure OSP-96-014. E0 Equipment Field
          Verification. The prerequisite in procedure OSP 96-014 for
          individuals performing the walkdowns was to read the procedure.
          The licensee qualified a number of individuals to perform the
          field walkdowns through a training program conducted in accordance
          with CP&L procedure TI-100. Conduct of Training. These
          individuals included Instrumentation and Control technicians.
          contract engineers, and personnel assigned to the E0 group who
          were qualified E0 engineers. The training for the qualified E0
          engineers consisted of reading the procedure. orientation and on-
          the-job training to become familiar with the walkdown and data
s        gathering process. For other personnel, the training included
          reading of the procedures, formal classroom lectures.
          demonstrations, performance of practical exercises, and on-the-job
          training. The walkdown group supervisor performed a detailed
          review of the result < of practical exercises and data gathered    <
          during initial walke is prior to signifying the individuals were
                                                              s. _        __
 
                                    34
  qualified to perform walkdowns. The training provided for the
  walkdcwn personnel exceeded the procedural requirements. The E0
  walkdown grou) supervisor stated that the level of training
  provided to t1e walkdown personnel war to assure that the walkdown
  results were very accurate and to preclude the need for repeat
  work. The inspectors revieweJ the training records for the
  walkdown personnel and verified that they had been trained in
  accordance with the licensee's program. The inspectors noted that
  the experience level for the walkdown personnel varied from a
  recent graduate engineer to individuals with more than 20 years of
  experience. The inspectors reviewed the walkdown inspection
  records prepared by various individuals in the walkdown group and
  noted that the original walkdown records were complete and
  accurate, with some exceptions. Discussions with the walkdown
  group supervisor disclosed that corrections noted on the records
  were the result of reviews perfnrmed to resolve discrepancies in
  the records. The changes were made as a result of additional
  walkdown inspections which were doc'mented in the records. In one
  case, an individual was terminated for failure to perform the
  walkdowns and complete the walkdown records properly.    This
  individual's work was reviewed by the licensee and corrected where
  necessary.
  The inspectors also reviewed the training and qualification
  records for E0 technical personnel. These records included
  previous work experience, education and training, and CP&L
  specific training applicable to the E0 project. This training
  included E0 technical reviewer, E0 design verifier E0
  calculations, and E0 ESR originator.    The inspectors also
  questioned the manager of the E0 group concerning work assignments
  within the E0 grou). That is, assignment of specific activities
  to individuals wit 1 previous experience in a particular area of
  specialization, such as review of requirements for qualification
  of motors or specific types of instrumentation. The E0 group
  manager has recently are)ared a directory of all engineers working
  within the E0 group w11c1 lists each engineer's experience and
  what work activities they have completed for the E0 project at
  Brunswick. The purpose of this directory was for the engineers
  within the group to know who has worked on various problems and
  issues so they could obtain assistance from these individuals when
  they become involved with similar technical issues. The directory
  was distributed'to personnel in the E0 group. The E0 group
  manager provided a copy of the directory to the inspectors and
  discussed the basis for the various work assignments within the
  group.which were based on the past work experience of the E0
  technical personnel,
c. Conclusions
  The inspector concluded that the licensee's program for training
  and qualification of E0 engineers meets NRC requirements.
                                                                      ,
 
    .
  .
                                                    35
        E8    Hiscellaneous Engineering Issues (37551, 92903)
        E8.1  (Closed) URI 50-325(324)/97-08-04: Control of Ecuioment Data Base
              System (EDBS) Information
              The licensee issued CR 97-02400. Non Validated EDBS Information,
              concerning rc, tine use of non-validated EDBS information. This wes
              associated with VIO 50 325(324)/97-08 03. Safety Relay Setting Change
              Made as Pen and Ink Changes to Procedure. The licensee replied to this
              violation on September 2. 1997. The reply discussed licensee corrective
'
              action regarding the use of EDBS. Likewise. the licensee responded on
              November 26. 1997 to VIO 50-325/97-11-01. Failure to Initiate Alternate
              Safe Shutdown Impairment. addressed corrective action fcr use of an EDBS
              non validated field for determination of an Alternate Safe Shutdown
              impairment.      Plant procedure OENP-33.6. Equipment Data Base System
              Control and Revision, provides instructions for control of EDBS
              information. Color coding of fields in the electronic database
              represent the various types of data present. This procedure provides
              direction that certain types of data are not to be used until verified.
              Accordingly two previous violations address the use of non-verified
              EDBS information. The licensee corrective actions for these violations
              are being completed. The requirements for the control of information
              are in procedure OENP-33.6. Previous items address the concern of this
              URI. therefore this item is closed.
        E8.2 (Closed) LER 50-325(324)/97-04:        Soent Fuel Shionina Cask Handlina
              Activities
              This report documented the discovery by the licensee that the heavy load
              analysis as described in tne UFSAR did not completely bound movement of
              the shiroing cask from the primary lift to the secondary lift with the
              valve box covers removed.      It was determined that movement of the cask
              with a non single failure proof yoke and less than full cask integr'ty
              constituted an unreviewed safety question (US0) in accordance with the
              requirements specified in 10 CFR Part 50.55        The failure to obtain
              prior approval for a previously unanalyzed condition was determined in
              IR 50-325(324)/97-12 to be a violation. In a letter to the NRC dated
              August 6. 1997, the licensee requested a license amendment for review of
              the US"    The licensee re evaluated findings relative to the 30 foot
              dro: ~cident and qualified the transfer yoke using guidance provided in
              NUR b 0612. Control of Heavy Loads at Nuclear Power Plants. This
              evaluation contended that a fuel shipping cask drop event was not
              credible. therefore operation with less than full cask integrity was no
              longer a problem due to acceptable redundancy in the lifting yoke. In a
              letter to the licensee dated December 2. 1997, the NRC accepted the
              licensee determination that operation with the valve covers removed
              would not compromise the health and safety of the public due to
              acceotable redundancy of the lift devices. Based on the acceptance by
              the NRC of the licensee's evaluation and issuance of the enforcement
              action as described in IR 50-325(324)/97-12 this item is closed.
      .-                        .                                  ._
 
    _ _ . . _ _ _ _ _ _ _ . . _ . _ . _ . _ _ _ .                                                        _ _ _ _ _ _ _ . _ _ _ . _ _
                    .
4
                                                                                            36
                          E8.3 (Closed) Inspector Followun item 50-325(324)/96-14-05.'Effect of EO
                                          Accuracy on Instrument Setooint Calculations.
'
                                            Review of procedures and various documents by an independent                                            *
                                            contJ1 tant had resulted in a number of questions regarding the
;
~
                                            effect of environmental effects (uncertainties) on instrument
                                            accuracy        The questions / concerns were documented in an E mail                                    *
                                          message dated June 20, 1997. subjert E0 and Instrument Accuracy,
                                            in order to address the issues raised in the June 20 E mail
                                            message,      a review of instrument setpoint calculations was
                                            performed by licensee instrumentation and controls (l&C)
                                            engineers. . The review was documented in ESR 97-000426, which was
'
                                            discussed in paragraph El.3. above. The inspectors also reviewed
                                            various instrument setpoint calculations (documented in paragraph
"
                                            E1.2. above) and determined that E0 accuracy has been aroperly
                                            considered in the instrument setpoint calculations. T1e
                                            ins)ectors had no further questions regarding instrument setpoint
                                            metloaology or accuracy at this time.
                          E8.4            JClosed) Violation item 50-325(324)/97-02-08. Failure to Imolement ao
                                              nsoection Procram for Safety-Related Miscellaneous Structural Steel
                                            The licensee responded to this violation in letters dated
                                                                                                                                                    *
                                          April 30. 1997, and June 26. 1997 Subject: Reply to Notice of
                                            Violation. The licensee's corrective actions included revision of
                                            Specification 248-107 and review of other specifications to assure OC
                                            inspection criteria required by applicable codes and standards
                                            referenced in the UFSAR had been included in the specifications.
                                            Specifications reviewed included the following: 248-117 - Installation
                                            of Piping Systems: 048 012 - Installation of Electrical Cables: 006 001
                                            - Design. Testing & Inspection of Concrete Mixes. Concrete Materials and
                                          High-Strength Bolts: 005-005 - Design. Testing, & Inspection of Concrete
                                          Mixes. Concrete Materials: 013 001 - Concrete Work: and 018-002.
                                          Miscellaneous Steel.              Additional corrective actions included inspection
                                            of a sample of safety related high strength bolts installed using
                                            Specification 248-107. The inspectors reviewed the results of the
                                            structural steel inspectior.s which were documented in ESR 97-00085.
-
                                            hiscellaneous Structural Steel Connection Inspections. The licensee
                                            also revised procedure MMP-013. to incorporate the specification 248-107
                                            changes and trained OC. engineering and planning personnel on the
                                            changes to specification 248-107 which now require additional QC
                                            inspections. The inspectors reviewed records which documented
                                            inspections performed for selected USl A-46 modifications completcd on
                                            Unit 1 during the Fall.1997 refueling outage and verified the
                                            structural steel inspections were completed in accordance with the
                                            revised procedures.
4
                          Ee.5 (Closed) Violation item 50-325(324)/97 08-07. Failure to initiate
                                            Condition Reports to [,0cument Nonconformina E0 Items
I
  .                                        The licensee
                                            September      2. 199 reshonded
                                                                          Subject: Reply  to this  violation
                                                                                              to Notice of Violation. in a letter    The dated
,
          _ _ ,                        _ . - . . _ . _ ,      . . . _ _          - , . -      _                                  -        -_ __,
 
                                        37
      licensee's corrective actions included training of E0 personnel on
      the corrective action program, a review of che E0 walkdown data
      sheets to identify any potential nonconforming conditions which
      had not been previously identified and dispositioned, and
      organizational changes to improve management o"ersight in the E0
      group. CR 97-02367 was initiated by the licensee on July 3. 1997
      to document and disposition the two s)ecific examples of failure
      to initiate CRs identified by NRC. Tie inspectors ceviewed the CR
      closecut records (CR was closed on December 14, 1997) and the
      licensee's corrective actions and verified that the actions were
      completed in accordance with the licensee's violation response.
                                IV. Plant SuppEt
R1    Radiological Protection and Chemistry Controls
RI.1 Use of locks to Control Access
  a. Insnection Stone (71750)
      The inspector verified a selected sampling of doors required to be
      locked, by plant TSs and procedures, fc r the purpose of radiation
      protection,
  b. Observations and Findinas
      The inspector reviewed Environmental & Radiological Control 0E&RC-0040.
      Control of Locked High Radiation and Very High Radiation Areas, to
      determine the controls used to lock high radiation area doors and
      barriers. The inspector located a sampling of the locked high radiation
      area doors specified in OE&RC-0040 and tested them to ensure that they
      were locked. The ins)ector found that all the locked high radiation
      doors tested were locced,
  c. Conclusions
      The ins)ector determined that each of the locked high radiation area
      dcors w11ch were checked were locked.    The ins)ector concluded that the
      licensee is satisfactorily controll1ng locked ligh radiation areas in
      the plant.
R1.2 Radioactive Material Controls
  a. insoection Scqoe (71750)
      The inspector conducted a housekeeping tour of radioactive material
      storage areas located in outside areas within the protected area,
  b, Observations and Findinas
      The inspector found several poor radiological work practices in the
      radiological material (RAM) storage area located aojacent to the
 
_      _ _ - - _ . _ _ _ _ _ . - _ - . - _ _ .      _            _ _ _ _ - _ _ _ _ _ . . _ _ _ _ . _ .
  .
                                                      38
                  Radiological Maintenance Service Building in the northwest corner of the
                  p.*otected area. A bucket containing scaffolding brackets was half
                  filled with water and was labeled as radioactive material. The label
                  identified the brackets as contaminated. This practice had the
                  possibility of allowing the potentially contaminated water to cause a
                  spread of contamination in an RAM storage area. There was also
                  scaffolding identified as radioactive lying unprotected on a wooden
                  pa l l e'. .
                  The ~icensee conducted a walkdown of this area and the radiological
                  service building, and identified multiple conditions requiring action.
                  These items were identified in CR 97-04122. Nonconforming Material
                  Condition,
      c.        Conclusions
                  The inspector determined that several poor radiological work practices
                  existed in a radioactive material storage area.
    S2            Status of Security Facilities and Equipment
    c2.1 Plant Access Control and Physical Barriers
      a.        Inspection Scone (71750)
                  The inspector verified the status and condition of the protected area
                  fencing,
      b.        Qbser"ations Jnd Findinas
                  The inspector performed a walkdown of the protected area fence. The
                  fence was inspected for integrity such as corrosion on the posts, gaps
                  in the fence, and general adequacy. The inepector noted no
                  deficiencies,
      c.        Conclusions
                  The inspector found the status and condition of the protected area fence
                  to be satisfactory.
    F1            Control of Fire Protection Activities
    F1.1 Operability of Fire Protection Facilities and Eauioment
      a.        Ipsoection Scone (64704)
                  The inspector reviewed the operation's fire protection daily impairment
                  reports on the facility's fire protection systems and features, and
                  inspected these items to determine the performance trends and the
                  material conditions of this equipment.
 
    .__ _ _ __.- _ _ _ _ _ _.__._                                                        _ _ _ _ . . _ -    -
                                                                                                                    _
                      4
4
.
:
'
                                                                    39                                              ,
                          b. -    Observations end Findinas                                                            !
                                  A review of the Loss Prevention Unit daily Impairment Reports for
                                -December 8 - 11, 1997.- indicated that the following fire-protection
,
                                  components or systems for safety related areas were out of service:                  ,
.
'
                                          fire Protection-System        ~ Number of Imoairments
                                          Thermo-Lag Fire Barriers              2
                                          Fire Doors                          6
                                                                                                                      '
,                                        Cable Coating-                        1
:                                    -
                                          Fire: Detection System -              3                                    1
                                          Fire Suppression System              3
                                  The inspector noted that a number of- fire doors were out of service.
                                  This high number was attributed to the current DG building fire door
                                  corrective action (door replacement and repairs) that was in process for
                                  discrepancies identified during a June 1997 licensee self assessment of
                                  the fire protection program.- Appropriate compensatory measures had been
-
                                -1mplemented for the fire protection features which were out of service.
                                  The impairment status report provided the licensee with a good means of
                                  identifying out-of-service fire protection equipment and provided status
-
                                  for compensatory measures that were implemented. The corrective
                                  maintenance on degraded fire protection systems was accomplished in a
                                                                                        -
                                  timely manner,
,
                                  During the plant tours the inspector noted that the maintenance and
                                  material condition of the fire protection equipment were satisfactory.
                            c.    Conclusions
'
                                  Correstive maintenance on degraded fire protection systems was
                                  accomplished in a. timely manner.>The maintenance and material condition
,
                                  of the fire protection equipment and features were satisfactory.
                                                                                    -
                                                                                                                      ,
                        F2        Status of Fire Protection Facilities and Equipment
                        F2.1 E3ssive Fire Barriers
                                                                                                                      .
                                  Fire barriers ~ include penetration seals. wraps, walls. structural member--
                                  fire resistanticoatings.. doors, dampers. etc. Fire barriers are used to
                                -prevent the spread of fire and to protect redundant safe shutdown
                                  equipment.    Laboratory testing of fire barrier materials is done only on
                                  a-limited range of test assemblies. In-)lant-installations can vary
                                                                                                                      -,
                                  from the tested configurations. -Under tie provisions of Generic Letter
                                  (GL) 86-10. Implementation of Fire Protection Requirements, licensees
                                  are permitted to develop engineering evaluations justifying such
                                  deviations.
                                                                                                                      ;
  w    -, ,, - . . .        -. - -                                                .    - - ,.              . -
 
                      . . _ _ _ _ _ - _ - _ _ - _ _ _ _ _ _ _ _ _ .
'
                                                                    40
  2.2  Silicone foam Penetration Seals
    a. Inspection Stone (64704)
      The inspector reviewed the fire barrier ,,ilicone foam penetration seal
      design end testing. The inspector compared as-built fire barrier
      silicone foam penetratioh seals to fire endurance test configurations to
      verify that the as-built penetration seals reviewed were qualified by
      appropriate fire endurance tests, representative of, and bounded by, the
      design and construction of the fire endurance test specimens. During
      plant walkdowns the inspector observed the installation configurations
      of selected fire barrier silicone foam 3enetration seals to unfirm that
      the licensee had established an accepta)le design basis for those fire
      barriers used to separate safe shutdown functions.
    b. Observations and Findinas
      The inspector reviewed the fire barrier seal design and testing for six
;      of ten fire barrier silicone foam seal penetrations, Additional reviews
I      are documented in NRC 1Rs 50-325(324)/92-31, 93 08. and 93-38.
      The inspector reviewed Brunswick Specification No. 118 003, Revision 7.
      Selection and Installation of Fire Barrier Penetration Seals: Corrective
      Maintenance Procedure OCMP-010, Revision 2, Installation of Fire
      Barrier, Pressure Boundary Penetration and Water / Moisture Seals: Fire
      Protection Procedure FFP-015. Revision 23, Fire Barrier Penetration Seal
      Work Control: Periodic Test OPT-34.6.7.12. Revision 3. Fire Barrier
      Penetration Seals: and the Fire Hazards Analysis (FHA) for the location
      and description of fire areas: and assessed the licensee's supporting
      technical justification and any available engineering evaluations for
      the sampled silicone foam type oenetration seals,
      The inspector's review focused on verifying that the following design
      and installation paramaters for the as-built configurations were
      adequately bounded and justified by the licensee's engineering
      evaluations:
      .      penetration opening sizes
      e      thermal mass of penetrating items
      e      clearances of penetrating items
      e      unexposed surface temperatures
      The insoector found that penetration seal field verification
      documentation was maintained by the licensee.                  However, the seal
      installers * qualification and training records were not readily
      available for review. Although the installation and repair procedures
      for penetration seals provided adequate guidance to ensure materials
      were installed per design requirements, the inspector could not verify
      that the established surveillance recuirements included vendor
      recommendations for inspection and icentification of silicone foam seal
      aging and shrinkage.
 
    - - - - - -                        _ _ _ - - - _ - - . - - - . - - -                                    - - - - . . - -
            '
  .                                                                                                                                :
                                                                          41
                      The licensee was unable to locate the penetration seal testing
                      documentation and the vtador data for the tested prototype
                      configurations or GL 8610 engineering evaluation documentation that
                      evahated the adequacy of the deviations from a tested fire barrier
                      contiguration. This does not satisfy the guidar.ce of GL 8610. The
i
                      licensee stated that industry documentation is available to support
                      silicone foam penetration seal installations at Brunswick but the
                      .tiformation was maintained at other Carolina Power and Light (CP&L)
                      sites.
                      The penetration seal testing documentation, vendor data and inspection
                      criteria, installer qualification and training records, and evaluations
                      of deviations from tested fire barrier configurations will be reviewed
                      during a subsequent NRC inspection. This is identified as IFl 50 325
                      (324)/97-13 04. Review of Licensee Records and Engineering Evaluations
                      to Establish the Fire Resistant Capabilities of Fire Rated Silicone foam
                      Penetration Seals,
                  c.  Conclusions
                      The inspector concluded that silicone foam penetration seal field
                      verification documentation was maintained by the licensee. The
                      installation and repair procedures for penetration seals provided
                      adequate guidance to ensure that materials were installed per design
                      requirements. However, the designs were not supported by seal testing
                      documentation, vendor data and inspection criteria, installer
                      qualification and training records, and engineering evaluations that
                      satisfy the guidance of GL 8610 for deviations from the fire barrier
                      configuration qualified by tests.
                F2.3 Fire Doors
                  a.  Insnection Scone (64704)
                      The inspector reviewed UFSAR Section 9.5.1.4.3.4.b. Fire Doors, and
                      performed plant walkdowns to verify that the UFSAR wording was
                      consistent with the observed plant installation configurations for
                      selected fire doors installed in fire barriers used to separate safe
                      shutdown functions.
                b.  Observations and Findinas
                      The UFSAR St.ction 9.5.1.4.3.4.b. Fire Doors, states that doors and
                      frames are either listed by a national testing laboratory or are
                      constructed similar to listed doors and frames. All doors and frames
                      have been evaluated to assure satisfactory ratings. Results are
                      documented in the FHA. During the review of the FHA the inspector
                      identified that, while evaluations of fire doors and frames existed. the
                    -licensee failed to document their results in the FHA. which is section
                      9.5.1.5 of the UFSAR.
                                                                                                                                  1
                                - - ,        r    ,,-                  ~    , - , , - , , - , - - - - - -            -  , -v ,
 
                          _ _ _ _ _ _ - _
                                                                          42
                      After discussions with the licensee. CR 97-04103 was issued to track the
l                        failure to provide the results of fire door evaluations in the FHA.
                        This UFSAR discrepancy was identified by the inspector and is discussed
                        in Section F2.4.
                      A review of the surveillance ins)ection and testing procedures for fire
                        doors was performed to confirm tlat the licensee specified fire door
                        clearance acce)tance criteria was in accordance with the guidance of
                        National Fire )rotection Association (NFPA) 80. Standard for Fire Doors
                        and Fire Windows. On December 10. 1997. the inspector observed ongoing
                        door replacement and repair activities for fire doors in the DG
                        building. No discrepancies were identified,
                    c.  Conclusions
I
                        The inspector concluded that fire door surveillance prc:edures and
                        acceptance criteria for verification o' fire daor clearances were in
                        accordance with NFPA quidance.                    Howevr a UFSAR discrepancy associated
                        documentation of fire door and frame eu.uations was identified.
                F2.4 UFSAR Review
                        A recent discovery of a licensee o)erating the facility in a manner
                        contrary to the UFSAR description lighlighted the need for a special
                        focused review that compares plant practices, procedures, and/or
                        parameters to the UFSAR descriptions. While performing the inspections
                        discussed in this report. the inspector reviewed the applicable portions
                        of the UFSAR that related to the areas inspected. The inspector
                        verified that the UFSAR wording was consistent with the observed plant
                        practices, procedures, and/or parameters.
                        The inspector reviewed UFSAR Section 9.5.1.4.3.4.b, Fire Doors, as part
                        of the fire protection program review activiti u , An inconsistency was
                        noted in that the licensee failed to document the results of evaluations
                        of fire doors and frames in the FHA which is section 9.5.1.5 of the
                        UFSAR.                This issue is discussed in Section F2.3. This item will be
                        identified as part of URI 50-325(324)/97-13-05. UFSAR Discrepancy Fire
                        Doors.
                F3      Fire Protection Procedures and Documentation
                F3.1 Fire Protection Procedures
                    a.    Insoection Scone (64704)
                        The inspector evaluated the adequacy and implementation of the
                        licensee s Eire Protection Program described in the UFSAR and in Plant
                        Operating Manual Fire Protection Procedure OPLP 01. Revision 6. Fire
                        Protection Program Document. In addition a comparison was made of the
                        program to selected NRC Safety Evaluation Reports which ap3 roved the
                        station fire protection program. The inspector reviewed t7e following
>
                        procedures for compliance with the NRC requirements and guidelines:
    .
  .  . . . - ..  ~    .
                                          ..
                                              .
                                                        .
                                                            . _ . . . . .
                                                                                                                ;
                                                                          .
                                                                                                .
                                                                                                    .
 
    .
  .
                                                  43
              -
                      OPLP-01. Revision 6. Fire Protection Program Document
              -
                      0FLP-01.1. Revision 12. Fire Protection Commitment Document
              -
                      OPLP-01.2 Revision 10. Fire Protection System Operability.
                      Action, and Surveillance Requirements
              -
                      FPP 005. Revision 15. Fire Watch Program
              -
                      FPP-008. Revision 24. Fire Protection Weekly inspection
              -
                      FPP 013. Revision 25. Transient Fire Load Evaluation
              -
                      FPP 014. Revision 17. Control of Combustible. Transient Fire loads
!
                      and Ignition Sources
              Plant tours were also performed to assess procedure complianc.e.
        b.    Obji.ervations and Findinas
              The listed procedures were issued to implement the facility's fire
              protection program.    These procedures contained requirements for program
              administration, controls over combust 1 oles arid ignition sources, fire
              watch duties and training, and operability requirements for fire
i
              protection systems and features. The 3rocedures were well written and
              met the licensee's commitments to the 1RC.
              General plant walkdown inspections were perfoimed by the inspector to
              verify: acceptable housekeeping; compliance with the ]lant's fire
              prevention procedures such as control of transient com)ustibles:
              operability of the fire detection and suppression systems: emergency              '
              lighting: and installation and operability of fire barriers, fire stop
              and penetration seals (fire doors, dampers, electrical penetration
              seals, etc.),
      c.      Conclusions
              General housekeeping was satisfactory. Fire retardant plastic sheeting
              and film materials were being used. Lubricants and oils were properly
                stored in approved safety containers. Controls for combustible gas bulk
                storage and cutting and welding operations were being enforced.
              Controls were being properly maintained for limiting t' alsient
                combustibles in designated separation zones and oth' restricted plant
            . areas.
      F5        Fire Protection Staff Training and Qualification
      F5.1 EireBrioade
        a.      Insoection Stone (64704)
,
                                          ,--                                            _____.
 
      _ _ _        _ _ _ . _ - . _ . _ _ _ _ _ . _ _ _ _ _ - - _ _ . - . . _ _ _ _ _ _                                                      __
                                                                                                                                                m
    .                                                                                                                                          1
                                                                                                                                                1
                                                                                                  44
                        The inspector reviewed the fire brigade organization and training
                        program for compliance with the NRC guidelines and program requirements.
                                                                                                                                                '
                b.      Observations and Findinos
'
                        The organization and training requirements for the plant fire brigade
                        were established by Fire Protection Procedure 0FPP-051. Loss Prevention
                        Emergency Response 0ualification/ Training and Drill Program. The fire
                        brigade for each of five shifts was composed of an operations support
                        fire protection technician shift incident commander (fire brigade
                        leader) and at least four additional brigade members consisting of
                        Auxiliary Operators. Chemistry Technicians and Maintenance personnel.
                        Each operations shift also had a Senior Reactor Operator / Reactor                                                      :
                        Operator Fire Brigade Advisor assigned to respond tr ' ires with the fire
                        brigade.
                        As of the date of this inspection, there were a total of 48 fire brigade
                        members 26 from operations and 22 from E&RC and Maintenance on the
                        pic t fire brigade. The inspector verified that sufficient shift
                        personal were available to staff each shift's fire brigade with at
                        least five qualified fire brigade members.
                        A review of the training records for the fire brigade members indicated
                        that the training, drill, respiratory and physical examination
                        requirements for each active member were up to date and met the
                        established site training requirements.
                        Fire Briaade Ecuioment:
                        The fire brigade turnout gear and a fire response vehicle and trailer
                        with fire brigade equi) ment was stored in the Operations / Fire Protection
                        equipment building. T1e_ inspector's inventory of the fire brigade
                        equipment indicated that a sufficient number of turnout gear, consisting
                        of coats, pants, boots, helmets, etc. , was provided to equip the fire
                        brigade members expected to respond in the event of a fire or other
                        emergency. The fire brigade turnout i., ear and fire fighting equipment
                        were being properly maintained,
              c.      Conclusions
                        The fire brigade organization and qualification training met the
                      -requirements of the site procedu.m                                            . Fire brigade turnout gear and fire
                        fighting eouipment were being properly maintained.
  __      .__    -
                                    - _ . -                  _          _              _ - - - -
                                                                                                        - _        _          ,  - _ _  --
                                                                                                                                            ..
 
                                                                                    l
e                                                                                  l
                                          45                                      j
  F6    Fire Protection Organization and Administration
  F6.1 Fire Protection Mananement and OraanizatioD
    a. Inspection Scope (64704),
        The licensee's management and administration of the facility's fire
        protection program were reviewed for compliance with the commitments to
        the NRC and to current NRC guidelines.
    b. Observations and Findinos
        During this report period the licensee reassigned the responsibility ior
        the administration and implementation of the fire protection program
        from the previous Loss Prevention Unit (LPU) to the Operations Shift and
        Support organizations. The LPU organization was dissolved.
        The designated onsite manager responsible for the administration and
        implementation of the fire protection program was the Operations
        Manager, This responsibility had been delegated to the Operations
        Support Superintendent. The Operations Support Superintendent was
        responsible for the station fire protection program, fire protection
        surveillance testing of fire protection systems and equipment, and
        ensuring that the aopropriate fire prevention procedures and fire
        b:'igade programs were implemented. A Fire Protection Program
        C0ordinator reported to the Operations Support Superintendent.
        Maintenarice of the 31 ant fire protection equipment was performed by the
        Maintenance Unit. cire protection related training was planned and
        conducted by the Brunswick Training Se: tion. Coordination of the
        station's fire protection program commitments and engineering functions
        was provided by a fire protection system engineer in the Brunswick
        Engineering Support Section,
    c. Conclusions
        The coordination and oversight of the facility's fire protection program
        had been reassigned from the previous LPU organization to Shift
        Operations. The new organizational structure met NRC guidelines and the
        licensee's fire protection program requirements.
  F7    Quality Assurance in Fire Protection Activities
  F7.1 Fire Protection Audits
    a. Insoection Scope (64704)
        The following audit report and the plant response to the issues were
        reviewed:
        -      Nuclear Assessment Section (NAS) Report B-FP-97-01. Brunswick Fire
              Protection and Loss Prevention Unit Assessment, dated
              August 1. 1997.
                                                                                  i
 
    ,
  .
                                              46
        b. Observations and Findinas
            The licensee's Nuclear Assessment Section performed an assessment of the
            fire protection program and LPU on June 16-27. 1997. The report for
            this assessment was Re) ort No. B FP-97 01. The assessment team
            determined that the LPJ fire prevention and fire response activities
            were adequate; however, its implementation of the fire protection
            )rogram was ineffective based on a number of program elements found to
            )e below acceptable standards. Findings from these assessments were
            categorized as strengths, issues, or weaknesses. The assessment report
            identified six program element issues and one weakness.
            The inspector reviewed the final audit report, the licensee's response
            to the identified issues. the planned corrective actions, and the NAS
            evaluation of the response adequacy.
            This NAS assessment of the facility's fire protection program was
            comprehensive and effective in identifying fire protection program
            performance deficiencies to management. The audit team identified
            deficiencies in LPU'c management oversight of fire protection
            procedures, training, problem identification, procedure performance
            standards, corrective actions, and personriel safety. Corrective actions
            in response to the identified issues were substantial and included a
            fire protection reorganization to integrate the former LPU organization
            into the shift Operations and Operations Sup) ort organizations under
            direct management of the Operations Support Manager,
        c. Conclusions
            The 1997 Nuclear Assessment Section assessment of tite facility's fire
            protection program was comprehensive and was effective in identifying
            fire protection program performance deficiencies to management. Planned
            corrective actions in response to the audit issues were substantial and
            included a fire protection reorganization.
                                  V.  Manaaetment Meetinas
      XI    Exit Meeting Summary
            The inspector presented the inspection results to members of licensee
            management at tN conclusion of the ins)ection on January 8,1998. Post
            inspection briefings were conducted on )ecember 12, 1997. The licensee
            acknowledged the findings presented. The licensee stated that they had
            not determined if clearance records were required QA records.
_
                                                                                    A
 
                    ..  __ _ _ _ _ _              _ _ _
                                                          - _ _ - _ - _ - _ _ - _ _ _ - - _
                                                                                              .
  .
.
                                                        47
                                      PARTIAL LIST OF PERSONS CONTACTED
    Licensee
    A. Brittain. Manager Security
    M. Christinziano, Manager Environmental and Radit lon Control
                                                                                            _
    W. Dorman. Supervisor Licensing and Regulatory Programs
    N. Gannon. Manager Maintenance
    J. Gawron. Manager Nuclear Assessment Section
    S. Hinnant. Vice President. Brunswick Steam Electric Plant
    K. Jury. Manager Regulatory Affairs
    R. Krich, Chief Engineer. Nuclear Engineering Department
    B. Lindgren. Manager Site Su) port Services
    J. Lyash. Manager Brunswick Engineering Support Section
    R. Mullis. Manager Operations
    Other licensee employees or contractors included office, operation,
    maintenance. chemistry, radiation, and corporate personnel.
                                        _
 
.
                                          48
                            INSPECTION PROCEDURES USED
  IP 37550:  Engineering
  IP 37551:  Onsite Eng11eering
  IP 61726.  Surveillance Observations
  IP 62700:  Maintenance Program implementation
  IP 62707:  Maintenance Observations
  IP 64704:  Fire Protection
  IP 71707:  Plant 0)erations
  IP 71714:  Freeze )rotection
  IP 71750:  Plant Support Activities
  IP 92700:  Onsite Followup of Written Reports of Nonroutine Events at Power
              Reactor Facilities
  IP 92901:  Followup - Operations
  IP 92902:  Followup - Maintenance
  IP 92903:  Followup - Engineering
                        ITEMS OPENED, CLOSED, AND DISCUSSED
  Opened
  50-325(324)/97-13-01    VIO  Failure to Retain TS Required QA Record (Section
                                07.2)
  50 325(324)/97 13-02    VIO  Inadequate Procedure for the Conduct of E0
                                Preventive Maintenance (Section M1.3, El.4.b.6)
  50 325/97-13-03          VIO  Failure to Note Abnormal TS Surveillance Values
                                (Section M3.1)
  50 325(324)/97-13-04    IFl  Review of Licensee Records and Engineering
                                Evaluations to Establish the Fire Resistant
                                Capabilities of Fire Rated Silicone foam
                                Penetration Seals (Section F2.2)
  50-325(324)/97-13-05    URI  UFSAR Discrepancy Fire Doors (Section F2.4)
  50 325(324)/97-13 06    IFl  Revisions to Procedure EGR-NGGC-0153 (Section
                                El.1)
  50-325(324)/97-13-07    IFl  Review Technical Evaluation of Terminal Block
                                Current Leakayc and the Effect on EQ Equipment.
                                (Section El.4.b.4)
  Closed
  50-325/96-15-01          URI  Vessel Disassembly Without Secondary Containment
                                (Section 08.1)
  50-325(324)/97 02-01    V10  Locked Valve Out of Position (Section 08.2)
  50-325/97 12 03          URI  Recirculation Pump Run back (Section 08.3)
                                                                                  )
 
      _ _ _ _ _ _ _ _ _ _
                .
    .
                                                              49
                          50-325(324)97-12-04  URI  Diesel Generator Low Voltage Auto Start Defeated
                                                      (Section 08.4)
                          50 325(324)/96-017-00 LER  Invalid Loss of Coolant Accident (Section M8.1)
                          50_-325/97_009-00    LER  Missed Increased Frequency inservice Testing
                                                      Requirement (Section M8.2)
                          50-325/97-001-00      LER  Rod Block Monitor Surveillance inadequacy
                                                      (Section M8.3)
                          50-325(324)/95-022 00 LER  High Pressure Coolant injection System Discharge
                                                      Flow Element Gasket Leak (Section M8.4)
'
                          50 325/97-05-02      IFl
                                                      Abnormal    CS Sp)arger Break Detector Indication
                                                      (Section Md.5
                          50 325/97-05-03      VIO  Inadequate CS Surveillance Procedure (Section
                                                      M8.5)
                          50 325/97-02          LER  Core Spray Header Differential Pressure
                                                      Instrumentation Inoperable (Section M8.5)
                          50-325(324)/97-02-04  VIO  Failure to implement Requirements of the
                                                      Maintenance Rule (Section M8.6)
                          50-325(324)/97-08-04  URI  Control of EDBS Information (Section E8.1)
                          50-325(324)/97-04    LER  Spent Fuel Shipping Cask Handling Activities
                                                      (Section E8.2)
                          50-325(324)/96-14-05  IFI  Effect of EQ Accuracy on Instrument Setpoint
                                                      Calculations (Section E8.3)
                          50-325(324)/97-02-08  VIO  Failure to Implement an Inspection Program for
                                                      Safety-Related Miscellaneous Structural Steel
                                                      (Section E8.4)
                          50-325(324)/97-08 07  VIO  Failure to Initiate Condition Reports to
                                                      Document Nonconforming EQ ltems (Section E8.5)
  r
                                    .
                                                    -
                                                                              .
                                                                                                        }
}}

Revision as of 07:30, 10 December 2024

Redirect to:

Retrieved from "https://kanterella.com/w/index.php?title=ML20199G750&oldid=4087446"