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                                                  U. S. NUCLFAR REGULATORY COMMISSION
.
                                                                REGION 11
U. S. NUCLFAR REGULATORY COMMISSION
                                Docket Nos:         50-325, 50 324
REGION 11
                                license Nos:       DPR 71. DPR 62
Docket Nos:
                                Report No:         50-325/97-13. 50-324/97 13
50-325, 50 324
                                Licensee:           Carolina Power & Light (CP&L)
license Nos:
                                Facility:           Brunswick Steam Electric Plant, Units 1 & 2
DPR 71. DPR 62
                                Location:           8470 River Road SE
Report No:
                                                      Southport, NC 28461
50-325/97-13. 50-324/97 13
                                Dates:               November 9 - December 27, 1997
Licensee:
                                  Inspectors:         C. Patterson Senior Resident Inspector
Carolina Power & Light (CP&L)
                                                      E. Brown Resident inspector
Facility:
                                                      G. Guthrie, inspector in Training
Brunswick Steam Electric Plant, Units 1 & 2
                                                      J. Coley Reactor inspector (M1.3. M8.6)
Location:
                                                      J. Lendhan. Reactor Inspector (E1.1. E1.4. E5.1. E8.3.
8470 River Road SE
                                                        E8.4. E8.5)
Southport, NC 28461
                                                      C. Doutt. Senior Instrumentation and Controls
Dates:
                                                        Engineer. Office of Nuclear Reactor Regulation
November 9 - December 27, 1997
                                                        (E1.1. E1.2. El.3)
Inspectors:
                                                      G. Wiseman. Reactor Inspection (F2.1. F2.2. F2.3,
C. Patterson Senior Resident Inspector
                                                        F3.1. F5.1 F6.1. F7.1)
E. Brown Resident inspector
                                  Approved by:       M. Shymlock. Chief. Projects Branch 4
G. Guthrie, inspector in Training
                                                      Division of Reactor Projects
J. Coley Reactor inspector (M1.3. M8.6)
                            9802040330 900123
J. Lendhan. Reactor Inspector (E1.1. E1.4. E5.1. E8.3.
                            PDR
E8.4. E8.5)
                            G    ADOCK 05000324
C. Doutt. Senior Instrumentation and Controls
                                              PDR
Engineer. Office of Nuclear Reactor Regulation
                                                                                                  Enclosure 2
(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
ADOCK 05000324
G
PDR
Enclosure 2
_ _ _ _ _ _ _ _ _ _ _ .


                                                                                    i
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                                                                                    \
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                                                                                    1
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                                      EXECUTIVE SUMMARY
EXECUTIVE SUMMARY
                        Brunswick Steam Electric Plant. Units 1 & 2
Brunswick Steam Electric Plant. Units 1 & 2
                    NRC Inspection Report 50 325/97 13. 50-324/97-13
NRC Inspection Report 50 325/97 13. 50-324/97-13
    This integrated inspection included aspects of licensee operations,
This integrated inspection included aspects of licensee operations,
    engineering, maintenance, and plant support. The report covers a 6-week
engineering, maintenance, and plant support.
    period of resident inspection; in addition. It includes the resu'ts of
The report covers a 6-week
    maintenance, engineering, and fire protection ir,pections by regional and
period of resident inspection; in addition. It includes the resu'ts of
    headquarters inspectors.
maintenance, engineering, and fire protection ir,pections by regional and
    Operations
headquarters inspectors.
    e    The inspector concluded that u.e cold weather program has been
Operations
          satisfactorily implemented. Adequate contingency plans and operator
The inspector concluded that u.e cold weather program has been
          checks for proper operation of the systems were noted in the procedures.
e
          Section 01.1).
satisfactorily implemented.
    *    The inspector concluded. from a safety system walkdown, that the
Adequate contingency plans and operator
          Containment Atmospheric Dilution system was being maintained as designed
checks for proper operation of the systems were noted in the procedures.
          (Section 02.1).
Section 01.1).
    *    The clearance reviewed was prepared. authorized, and implemented in
The inspector concluded. from a safety system walkdown, that the
          accordance with procedure (Section 02.2),
*
    e    The inspector concluded that the Plant Nuclear Safety Committee meeting
Containment Atmospheric Dilution system was being maintained as designed
          provided an effective review of Unit I readiness for restart (Section
(Section 02.1).
          07.1).
The clearance reviewed was prepared. authorized, and implemented in
    e      Inspe.; tor review determined that clearance records were not retained in
*
          accorcance with Technical Specifications (TS). The failure to maintain
accordance with procedure (Section 02.2),
          clearance records in accordance with TS was a violation (Section 07.2).
The inspector concluded that the Plant Nuclear Safety Committee meeting
    *    The control of a short duration mid-cycle o:tage was excellent (Section
e
          07.3).
provided an effective review of Unit I readiness for restart (Section
    *    Licensee investigation determined that removal of the IB Reactor
07.1).
          feedwater Pump at too high a power level caused larger than expected
Inspe.; tor review determined that clearance records were not retained in
          level transients. These transients combined with the improper
e
          functioning of the level contacts in the Reactor Recirculation Run back
accorcance with Technical Specifications (TS). The failure to maintain
          logic circuitry, resulted in the November 5-6. 199/ run backs (Section
clearance records in accordance with TS was a violation (Section 07.2).
          08.3).
The control of a short duration mid-cycle o:tage was excellent (Section
    *    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
07.3).
          oberabilityconditionsduringtheimplementationoftt.eirclearance
Licensee investigation determined that removal of the IB Reactor
          ( ection 08.4).
*
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 *
F
                                                                                      1
*
                                                2
2
      Maintenance
Maintenance
      e    Movement of the spent fuel shi) ping cask was perforrxo in accordance
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.
e
            Adequate supervisory oversight was present during movement of the cask
with methodology approved by t1e NRC in a letter dated December 2, 1997.
            (Section M1.1).
Adequate supervisory oversight was present during movement of the cask
      *    The inspector observed performance of calibration of two Reactor Core
(Section M1.1).
            Isolation Cooling (RCIC) pressure switches. The work activities were
The inspector observed performance of calibration of two Reactor Core
            completed without any identified questions or concerns (Section M1.2).
*
      *    Maintenance activities observed relating to equipmert qualification of
Isolation Cooling (RCIC) pressure switches.
            electrical equipment were found to be conducted in a thorough and
The work activities were
            effective manner (Section M1.3).
completed without any identified questions or concerns (Section M1.2).
      .    A violation was identified for a preventive maintenance procedure not
Maintenance activities observed relating to equipmert qualification of
            indicating specific E0 requirements. This omission resulted in
*
            deficient Nelson flame seals in motor control centers not being detected
electrical equipment were found to be conducted in a thorough and
            during scheduled preventive maintenance activities (Section M1.3).
effective manner (Section M1.3).
      *    The licensee continues to struggle with proper dispositioning of
A violation was identified for a preventive maintenance procedure not
            abnormal indications. The failure to maintain the Daily Surveillance
.
            Report in accordance with procedure was a violation. Abnormal values
indicating specific E0 requirements.
            observed fer the Steam Jet Air Ejector radiation monitor and subsequent
This omission resulted in
            test indicated potential fuel failure for Unit 1 (Section M3.1),
deficient Nelson flame seals in motor control centers not being detected
      *    The licensee identified that the Unit 2 Core Spiay sparger differential
during scheduled preventive maintenance activities (Section M1.3).
            alarm setpoints were outside of the TS allowable range. The cauce was
The licensee continues to struggle with proper dispositioning of
            attributed to voiding of the sparger nozzles similar to the phenomenon
*
            identified previously on Unit 1. The alarm setpoints were adjusted and
abnormal indications.
            the associated documentation was updated (Section M8.5).
The failure to maintain the Daily Surveillance
      -Engineerino                                                                   >
Report in accordance with procedure was a violation.
      +    An additional example of a violation was identified for an inadequate
Abnormal values
            procedure for the conduct of E0 maintenance (Section E1.4).     Two
observed fer the Steam Jet Air Ejector radiation monitor and subsequent
            inspector followup items were identified to review revisions to
test indicated potential fuel failure for Unit 1 (Section M3.1),
            instrument setpoint procedures and to review terminal block leakage
The licensee identified that the Unit 2 Core Spiay sparger differential
            current evaluations (Section El.1 and Section E1.4).
*
      *    A weakness was identified regarding a procedure reference to a drawing
alarm setpoints were outside of the TS allowable range.
            for accident temperature data which was not available for use and
The cauce was
            wording inconsistencies in the procedure (Section E1.1).
attributed to voiding of the sparger nozzles similar to the phenomenon
      *    The licensee was making progress in resolution of the technical issues
identified previously on Unit 1.
            and closure of CRs and JCOs (Section E1.4). The licensee training and
The alarm setpoints were adjusted and
            qualification for E0 personnel meets NRC requirements (Section E5.1).
the associated documentation was updated (Section M8.5).
              Instrument setpoint calculations were technica ly adequate and complied
-Engineerino
            with NRC requirements (Section E1.2).
>
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
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                                                            _ _ _ _ _ _ _ _ _-
_ _ _ _ _ _ _ _
      .
_-
    .
.
                                                  3
.
        Plant Support
3
        .      The ins)ector determined that each of the locked high radiation area
Plant Support
                doors w11ch were checked were locked.   lhe ins)ector concluded that the
The ins)ector determined that each of the locked high radiation area
                licensee is satisfactorily controlling locked ligh radiation areas in
.
                the plant (Section Rl.1).
doors w11ch were checked were locked.
        .
lhe ins)ector concluded that the
                The inspector determined that several poor radiological work practices
licensee is satisfactorily controlling locked ligh radiation areas in
                existed in a radioactive material storage area (Section Rl.2).
the plant (Section Rl.1).
        *
The inspector determined that several poor radiological work practices
                The inspector found the status and condition of the protected area fence
.
i               to be satisfactory (Section S2.1).
existed in a radioactive material storage area (Section Rl.2).
        .
The inspector found the status and condition of the protected area fence
                Corrective maintenance on degraded fire protection systems was
*
i
to be satisfactory (Section S2.1).
Corrective maintenance on degraded fire protection systems was
.
accomplished in a timely manner.
The maintenance and material condition
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!
                accomplished in a timely manner. The maintenance and material condition
of the fire protection equipment and features were satisfactory
                of the fire protection equipment and features were satisfactory
(Section F1.1).
                (Section F1.1).
The inspector concluded that silicone foam penetration seal field
        .      The inspector concluded that silicone foam penetration seal field
.
                verificction documentation was maintained by the licensee. The
verificction documentation was maintained by the licensee.
                inst 311ation and repair procedures for penetration seals provided
The
                adequate guidance to ensure that materials were installed per design
inst 311ation and repair procedures for penetration seals provided
                requirements. However, the designs were not supported by seal testing
adequate guidance to ensure that materials were installed per design
                documentation, vendor data and inspection criteria, installer
requirements.
                qualification and training records, and engineering evaluations that
However, the designs were not supported by seal testing
                satisfy the guidance of Generic Letter 86-10 for deviations from the
documentation, vendor data and inspection criteria, installer
                fire barrier configuration qualified by tests (Section F2.2).
qualification and training records, and engineering evaluations that
        .
satisfy the guidance of Generic Letter 86-10 for deviations from the
                The inspector concluded that fire door surveillance procedures and
fire barrier configuration qualified by tests (Section F2.2).
                acceptance criteria for verification of fire door clearances were in
The inspector concluded that fire door surveillance procedures and
                accordance with National Fire Protection Association (NFPA) guidance.
.
                However, an updated Final Safety Analysis Report (UFSAR) discrepancy
acceptance criteria for verification of fire door clearances were in
                associated documentation of fire door and frame evaluations was
accordance with National Fire Protection Association (NFPA) guidance.
                  identified (Section F2.3).
However, an updated Final Safety Analysis Report (UFSAR) discrepancy
        .      General housekeeping was satisfactory. Fire retardant plast.ic sheating
associated documentation of fire door and frame evaluations was
                and film materials were being used. Lubricants and oils were properly
identified (Section F2.3).
                stored in approved safety containers. Controls for combustible gas bulk
General housekeeping was satisfactory.
                storage and cutting and welding operations were being enforced.
Fire retardant plast.ic sheating
                Controls were being properly maintained for limiting transient
.
                combustibles in designated separation zones and other restricted plant
and film materials were being used.
                areas (Section F3.1).
Lubricants and oils were properly
        .
stored in approved safety containers.
                The fire brigade organization and qualification training .act the
Controls for combustible gas bulk
                requirements of the site Procedures. Fire brigade turnout gear and fire
storage and cutting and welding operations were being enforced.
                  fighting equipment were being properly maintained (Section F5.1).
Controls were being properly maintained for limiting transient
        .
combustibles in designated separation zones and other restricted plant
                The coordination and oversight of the tacility's fire protection program
areas (Section F3.1).
                had been reassigned from the previous Loss Prevention Unit organization
The fire brigade organization and qualification training .act the
                to shift. Operations. The new organizat.onal structure met NRC
.
                guidelines and the licensee's fire protection program requirements
requirements of the site Procedures.
                  (Section F6.1).                                                          .
Fire brigade turnout gear and fire
                                                                                          l
fighting equipment were being properly maintained (Section F5.1).
  1
The coordination and oversight of the tacility's fire protection program
9
.
      .
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
4
  corrective actions in response tc the audit issues were substantial and
The 1997 Nuclear Assessment Section assessment of the facility's fire
  included a fire p.'otection reorganization (Section F7.1).
.
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
ReDort Details
    ~ Summary of Plant Status
~ Summary of Plant Status
    Unit I returned to power o)eration on November 14. 1997, following a mid-cycle
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
outage that began on Novem)er 5. 1997, to remove leaking fuel assemblies.
    leaking fsel assemblies were identified and removed during the mid cycle
Two
    outage.     However, indications of a potential fuel leaker remained after the
leaking fsel assemblies were identified and removed during the mid cycle
    unit returned to full power operation. At the end of the report period the
outage.
    unit had been on-line 42 days.
However, indications of a potential fuel leaker remained after the
    Unit 2 operated continuously during this report period. At the end of the
unit returned to full power operation.
    report period the unit had been on-line continuously for 59 days.
At the end of the report period the
    Due to concerns about the control room dose, the licensee imposed an
unit had been on-line 42 days.
    administrative limit on lodine until a Technical Specification (TS) amendment
Unit 2 operated continuously during this report period. At the end of the
    submitted was a) proved.   The licensee made a orocedure change to
report period the unit had been on-line continuously for 59 days.
    Administrative procedure 0Al-81. Water Chemistry Guidelines, setting the limit
Due to concerns about the control room dose, the licensee imposed an
    at 0.1 microcurie per gram dose equivalent L 'ine 131 compared to the TS value
administrative limit on lodine until a Technical Specification (TS) amendment
    of 0.2 microcurie per gram. Also, the licet ;e has been providing weekly
submitted was a) proved.
    water chemistry data to NRR and the Resident Inspector for review. None of
The licensee made a orocedure change to
    the data reviewed has exceeded the administrative limit.
Administrative procedure 0Al-81. Water Chemistry Guidelines, setting the limit
    Due to a reconstitution of the Environmental Qualification (EO) program and
at 0.1 microcurie per gram dose equivalent L 'ine 131 compared to the TS value
    items identified, there are 12 of 24 Justification for Continued Operation
of 0.2 microcurie per gram.
    (JCO) that remain open for both units. The following provides the status of
Also, the licet ;e has been providing weekly
    the EQ JCOs and associated Engineering Service Requests (ESRs):
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
I
            Closed
Closed
.
.
'
'
1)
ESR 97-00087. E0-Type JC0 for Improperly Configured Conduit Seal.
!
!
              1)   ESR 97-00087. E0-Type JC0 for Improperly Configured Conduit Seal.
2)
              2)  ESR 97-00574 Greyboot Connectors.
ESR 97-00574 Greyboot Connectors.
              3)   ESR 97-00329 (old ESR 96-00625). E0 Type JC0 for EQ Fuses Without
3)
                  a Qualification Data Package (00P).
ESR 97-00329 (old ESR 96-00625). E0 Type JC0 for EQ Fuses Without
              4)   ESR 97-00289. Post A cident Sampling System (PASS) Valve Limit
a Qualification Data Package (00P).
                  Switch Panel Wiring.
4)
              5)   ESR 97-00238. JC0 for Standby Gas Treatment Motor Operated Valve
ESR 97-00289. Post A cident Sampling System (PASS) Valve Limit
                  (MOV) Position Indicator Rheostat.
Switch Panel Wiring.
              6)   ESR-97-00534. GE c'                   Type Terminal Strips.
5)
              7)   ESR 97-00513. In-b                     Drywell Electrical Penetrations.
ESR 97-00238. JC0 for Standby Gas Treatment Motor Operated Valve
              8)   ESR 97-00535. Target Rock Solenoids TB Spray.
(MOV) Position Indicator Rheostat.
              9)   ESR 97-00449, Degraded Junction Boxes.
6)
            19)   ESR 97-00250. Conduit Union in EQ Boundary.
ESR-97-00534. GE c'
            11)   ESR 96-00425. Evaluation of E0 sealants.
Type Terminal Strips.
            12)   ESR 97-00523. High Pressure Coolant Injection Auxiliary Oil Pump
7)
                  Motor Unit 1.
ESR 97-00513. In-b
            0P10
Drywell Electrical Penetrations.
            13)   ESR 97-00446. GE Radiation Detectors. closure date to be
8)
                  determined (TBD).
ESR 97-00535. Target Rock Solenoids TB Spray.
            14)   ESR 96-00503. Associated Circuit E0. closure date TBD.
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.
.
.
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-
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.
cycle outage.
                                            I. Ooerations
Unit 2 o)erated continuously; however there were 12
    01   Conduct of Operaticns
outstanding JCOs in the E0 area for both units.
    01.1 Cold Weather Preparation
I. Ooerations
    a.   Insoection Scone (71714)
01
          The inspector reviewed the licensee's cold weather program to determine
Conduct of Operaticns
          whether it had been effectively implemented.
01.1 Cold Weather Preparation
    b.   Observations and Findinas
a.
          The inspector reviewed the licensee's cold weather 3rogram for adequacy
Insoection Scone (71714)
          and implementation by reviewing their Cold Weather 3111 and Freeze
The inspector reviewed the licensee's cold weather program to determine
          Protection Procedure. Operating Instruction 001-01.02: Fire Protection
whether it had been effectively implemented.
          Procedure 0FPP-024. Freeze Protection of Fire Suppression System; and
b.
          Preventive Maintenance Procedure OPM-HT001. Preventive Maintenance on
Observations and Findinas
          Plant Freeze Protection and Heat Tracing. The inspector determined that
The inspector reviewed the licensee's cold weather 3rogram for adequacy
          the procedures were adequately implemented. Additionally, the
and implementation by reviewing their Cold Weather 3111 and Freeze
          procedures were adequately employed on multiple cold weather days. as
Protection Procedure. Operating Instruction 001-01.02: Fire Protection
          observed by the inspector.
Procedure 0FPP-024. Freeze Protection of Fire Suppression System; and
          The inspector conducted a walkdown of plant syn. , which were exposed
Preventive Maintenance Procedure OPM-HT001. Preventive Maintenance on
          to cold weather. Systems which were heat traced were observed for
Plant Freeze Protection and Heat Tracing. The inspector determined that
          adequacy. The inspector looked for systems that did not have cold
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
3
        by any procedure. The licensee stated that this heat trace system was
weather. heat trace installed.
        being controlled b," operator knowledge only. The licensee initiated a
The inspector determined that the
        procedure change request to place this heat trace system into their cold
operation of the Makeup Water Tank system heat trace was not controlled
        weather procedures. The inspector noted on the Unit 2 Condensate
by any procedure.
        Storage Tank. High Pressure Coolant Injection (HPCI)/ Reactor Core
The licensee stated that this heat trace system was
        Isolation ~ Cooling (RCIC) level switch vent line that a six inch portion
being controlled b," operator knowledge only. The licensee initiated a
        of the lagging was missing at the top of the vent line and that the tin
procedure change request to place this heat trace system into their cold
        shielding was missing around the lagging at an elbow on the vent line.
weather procedures.
        The lagging was wetted and degraded at the elbow. The inspector
The inspector noted on the Unit 2 Condensate
        discussed these two items with the licensee. The licensee did not
Storage Tank. High Pressure Coolant Injection (HPCI)/ Reactor Core
        warrant these deficiencies as requiring corrective action. The
Isolation ~ Cooling (RCIC) level switch vent line that a six inch portion
        inspector did not find other systems requiring heat trace that were not
of the lagging was missing at the top of the vent line and that the tin
        heat traced based on present system conditions and projected use of the
shielding was missing around the lagging at an elbow on the vent line.
        systems observed.
The lagging was wetted and degraded at the elbow.
  c.   Conclusions
The inspector
        The inspector concluded that the cold weather program has been
discussed these two items with the licensee.
        satisfactorily implemented. Adequate contingency plans and operator
The licensee did not
        checks for proper operation of the systems were noted in the procedures.
warrant these deficiencies as requiring corrective action.
  02   Operational Status of Facilities and Equipment
The
  02.1 Containment Atmosoheric Dilution (CAD) System Walkdown
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)
a.
        On December 10. 1997, the inspector performed a walkdown of the CAD
Insoection Scope (71707)
        system in the Nitrogen and Off-Gas Services Building.
On December 10. 1997, the inspector performed a walkdown of the CAD
  b.   Observations and Findinos
system in the Nitrogen and Off-Gas Services Building.
        The CAD system is described in Updated Final Safety Analysis Report
b.
        (UFSAR) Section 6.2.5. Combustible Gas Control in Containment. The CAD
Observations and Findinos
        system provides long-term nitrogen makeup after a Loss of Coolant
The CAD system is described in Updated Final Safety Analysis Report
        Accident (LOCA).   This function is accomplished by vaporizing liquid
(UFSAR) Section 6.2.5. Combustible Gas Control in Containment.
        nitrogen and feeding it into containment as required to maintain an
The CAD
        oxygen concentration at or below five percent. The system is designed
system provides long-term nitrogen makeup after a Loss of Coolant
        to Engineered Safety Feature (ESF) standards, all equipment for CAD
Accident (LOCA).
        service is designed with suitable redundancy and interconnections such
This function is accomplished by vaporizing liquid
        that no single failure of an active component will render the system
nitrogen and feeding it into containment as required to maintain an
        inoperable. This equipment includes one liquid nitrogen storage vessel.
oxygen concentration at or below five percent.
        two electric vaporizers, two flow-regulating stations. flow and
The system is designed
        temperature indicators. and appropriate redundant valves and
to Engineered Safety Feature (ESF) standards, all equipment for CAD
        interconnecting piping.
service is designed with suitable redundancy and interconnections such
        The inspector traced the system piping in the Nitrogen and Off-Gas
that no single failure of an active component will render the system
        Services Building. The configuration was compared to plant drawing
inoperable.
        0 02560. Containment Atmospheric Control System. The configuration was
This equipment includes one liquid nitrogen storage vessel.
        found to be like the plant drawing. The inspector observed an inch of
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
4
          frost on the outside of the piping insulation on both sides of valve
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
HV-11.
        an 85 pound pressure regulating valve.
This valve is a manual isolation between the nitrogen tank and
        The inspector questioned why the frost was on the line. The licensee
an 85 pound pressure regulating valve.
        stated that the 90 pound relief valve setpoint was near the controlling
The inspector questioned why the frost was on the line.
        pressure of the 85 pound regulator and some nitrogen was venting off.
The licensee
        The redundant pressure regulating valve was isolated and it's isolation
stated that the 90 pound relief valve setpoint was near the controlling
        valve (HV-12) was closed. The inspector questioned by keeping HV-12
pressure of the 85 pound regulator and some nitrogen was venting off.
        closed, if the system was single failure proof. The licensee initiated
The redundant pressure regulating valve was isolated and it's isolation
        CR 97-04128. CAD Tank Isolation Valve, to address this issue, The
valve (HV-12) was closed. The inspector questioned by keeping HV-12
          licensee concluded that no automatic action was required to address a
closed, if the system was single failure proof.
        LOCA.   Manual alignment of the pressure regulator was acceptable since
The licensee initiated
        this was a long term post-LOCA action.
CR 97-04128. CAD Tank Isolation Valve, to address this issue,
    c. Conclusions
The
        The inspector concluded, from a safety system walkdown, that the CAD
licensee concluded that no automatic action was required to address a
          system was being maintained as designed.
LOCA.
    02.2 Clearance Verification
Manual alignment of the pressure regulator was acceptable since
l     a. Insoection Scoce (71707)
this was a long term post-LOCA action.
        The inspector reviewed the tagout for the Unit 2 Residual Heat Removal
c.
                        -
Conclusions
          (RHR) system to verify proper clearance preparation, authori7     n. and
The inspector concluded, from a safety system walkdown, that the CAD
          implementation,
system was being maintained as designed.
      b. Observations and Findinas
02.2 Clearance Verification
          On December 10. 1997, the inspector performed verification of the proper
l
          alignment and tagging of clearance 2-97-1781 on the Unit 2 RHR System.
a.
          All accessible components were verified to-be in the proper position
Insoection Scoce (71707)
        with the appropriate tags in place. The inspector reviewed Nuclear
The inspector reviewed the tagout for the Unit 2 Residual Heat Removal
          Generation Group Standard Procedure OPS-NGGC-1301. Equipment Clearance.
-
          The clearance package was adequately prepared, authorized, aad
(RHR) system to verify proper clearance preparation, authori7
          implemer.ted. The inspector subsequently verified proper clearance
n. and
          removal for those accessible components.
implementation,
      c. Conclusions
b.
          The clearance reviewed was prepared, authorized, and implemented in
Observations and Findinas
          accordance with procedure,
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
i
  w
w


                                    . _ . _ _ _ _ _ _
. _ . _ _ _ _ _ _
                                                                              . .   . .. ._ .
. .
                                                                                                  .
. ..
                                                          5
._
  07     Quality Assurarm in Operations
.
  07.1 Restart Plant Nuclear Safety Committee (PNSC)
.
    a. Insoection Scone (71707)
5
        On November 11 and 12. 1997, the inspector attended the Unit 1 PNSC
07
        restart assessment following a mid-cycle outage to replace two leaking
Quality Assurarm in Operations
        fuel assemblies,
07.1 Restart Plant Nuclear Safety Committee (PNSC)
    b. Observations and Findinos
a.
        On November 11, 1997. PNSC was convened to review Unit I readiness for
Insoection Scone (71707)
        restart. The committee reviewed the fuel sipping results and core
On November 11 and 12. 1997, the inspector attended the Unit 1 PNSC
        reload.   Other maintenance activities during the outage were also
restart assessment following a mid-cycle outage to replace two leaking
        reviewed.
fuel assemblies,
        The meeting was conducted in accordance with TS with attendance by all
b.
        primary members, with no alternates. The meeting provided a thorough
Observations and Findinos
        discussion of all agenda items. The PNSC Chairman concluded that the
On November 11, 1997. PNSC was convened to review Unit I readiness for
        discussion of recirculation pump runbacks that occurred on November 5.
restart. The committee reviewed the fuel sipping results and core
        1997, during removal of the reactor feed pumps during the planned
reload.
        shutdown was not complete.                   This item was statused as a restart
Other maintenance activities during the outage were also
        constraint requiring another PNSC review prior to restart. Noteworthy
reviewed.
        in the review was the risk assessment review conducted for a failed
The meeting was conducted in accordance with TS with attendance by all
        Control Rod Drive (CRD) pump. During the mid-cycle outage one of the
primary members, with no alternates.
        two CRD pump motors failed. The Probabilistic Safety Analysis (PSA)
The meeting provided a thorough
        person attended the comnittee meeting and presented the results from
discussion of all agenda items.
        running the risk assessment model considering failure of both CRD Jumps.
The PNSC Chairman concluded that the
        This risk was determined acceptable based on other TS required higi
discussion of recirculation pump runbacks that occurred on November 5.
        pressure injection sources such as HPCI and RCIC.
1997, during removal of the reactor feed pumps during the planned
        On November 12. 1997, the inspector attended a second meeting. In this
shutdown was not complete.
        meeting discussion was held regarding the problem with run backs and it
This item was statused as a restart
        was concluded that this was due to a design deficiency that was already
constraint requiring another PNSC review prior to restart.
        corrected and installed on Unit 2 and scheduled for Unit 1 at the time
Noteworthy
        of the next refueling outage,
in the review was the risk assessment review conducted for a failed
    c. Conclusions
Control Rod Drive (CRD) pump.
        The inspector concluded that the PNSC meeting provided an effective
During the mid-cycle outage one of the
        -review of Unit I readiness for restart.
two CRD pump motors failed. The Probabilistic Safety Analysis (PSA)
  07.2 Retention of Clearance Records
person attended the comnittee meeting and presented the results from
    a.   Insoection Scope (71707)
running the risk assessment model considering failure of both CRD
\
Jumps.
        The inspector reviewed whether configuration management documents,
This risk was determined acceptable based on other TS required higi
        specifically ciearances, were retained in accordance with TS 6.10. This
pressure injection sources such as HPCI and RCIC.
        specification requires that facility records be retained in accordance
On November 12. 1997, the inspector attended a second meeting.
        with the American National Standards Institute (ANSI) N45.2.9-1974
In this
        Collection. Storage, and Maintenance of Quality Assurance Records.
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-               -
6
      cooler. Tha clearance. 2-97-1531. was hung _resulting in a configuration
b.
      change for the B recirculation pump, but no maintenance on the system
Observations and Findinas
      was performed. The clearance was removed from the field, thus restoring
During ins)ector review of clearance errors which resulted in damage to
      the system, and " rolled back" to allow use at a later date.
the Unit 23 recirculation pump seals, the licensee was unable to locate
      Subsequently, a scheduler requested the clearance be deleted due to the
a clearance hung to facilitate repairs on the recirculation motor oil-
      repair activities being complete and approved without need for the
-
      clearance boundary. As a result of the deletion of the clearance, no
cooler. Tha clearance. 2-97-1531. was hung _resulting in a configuration
      record of the change in plant configuration was retained.
change for the B recirculation pump, but no maintenance on the system
      The inspectoi : viewed TS 6.10. UFSAR Section 1.8. Regulatory Guide
was performed.
      1,88, and ANS1 N45.2.9-1974.                     fhe inspector questioned the correctness
The clearance was removed from the field, thus restoring
      of not retaining the clearance. Since a configuration change did occur
the system, and " rolled back" to allow use at a later date.
      despite the recirculation motor cooler activities not needing the cooler
Subsequently, a scheduler requested the clearance be deleted due to the
      isolated. Nuclear Records Management Procedure ORMP-001. Indexing of
repair activities being complete and approved without need for the
      Plant Records. defined those records required to be retained to satisfy
clearance boundary. As a result of the deletion of the clearance, no
      the 0A requirements stated in ANSI N45.2.9-1974. Discussion with the
record of the change in plant configuration was retained.
      licensee revealed that the records required to be retained did not
The inspectoi
      include clearances. The inspector reviewed the Nuclear Generation Grou)
: viewed TS 6.10. UFSAR Section 1.8. Regulatory Guide
      Standard Procedure OPS-NGGC-1301. Equipment Clearance, and the Brunswicc
1,88, and ANS1 N45.2.9-1974.
      Required Records List.                 Neither document required that clearances be
fhe inspector questioned the correctness
      retained.
of not retaining the clearance. Since a configuration change did occur
      TS 6.10 requires facility records shall be maintained in accordance with
despite the recirculation motor cooler activities not needing the cooler
      ANSI N45.2.9-1974. ANSI N45.2.9-1974, in Section 3.2.7. Retention of
isolated.
      Records. states that Appendix A to the standard defined the types of 0A
Nuclear Records Management Procedure ORMP-001. Indexing of
      records and the recommended retention periods. The failure to maintain
Plant Records. defined those records required to be retained to satisfy
      data sheets or logs on equipment alignment consistent with ANSI N45.2.9-
the 0A requirements stated in ANSI N45.2.9-1974.
      1974 is a violation. This violation is identified as VIO 50-325
Discussion with the
      (324)/97-13-01. Failure to Retain TS Required-0A Record.
licensee revealed that the records required to be retained did not
  c. Conclusion
include clearances. The inspector reviewed the Nuclear Generation Grou)
      Inspector review determined that clearance records were not retained in
Standard Procedure OPS-NGGC-1301. Equipment Clearance, and the Brunswicc
      accordance with TS. The failure to maintain clearance records in
Required Records List.
      accordance with TS was a violation.
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)
07.3 Mid-Cycle Outaae (71707)
  a. Insoection Scope
a.
      The inspector reviewed the mid-cycle outage activities to remove the
Insoection Scope
      leaking fuel assemblies.
The inspector reviewed the mid-cycle outage activities to remove the
  b. Observations and Findinas
leaking fuel assemblies.
      Unit 1 was returned to power operation on November 14. 1997. This
b.
      completed a mid-cycle outage in eight days. The unit was shutdown.
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
7
          record. T11s was accomplished with plant personnel without any major
leaking fuel assemblies identified, removed, fuel reloaded and returned
          problems. This outage was planned and controlled similar to a regular
to power o)eration.
          refueling outage.
This short duration outage was the quickest on
      c. Conclusions
record.
          The control of a short duration mid-cycle outage was excellent.
T11s was accomplished with plant personnel without any major
    08   Miscellaneous Operations Issues (92700, 92901)
problems.
    08.1 (Closed) Unresolved Item (URI) 50-325/96-15-01:                       Vessel Disassembly
This outage was planned and controlled similar to a regular
          Without Secondary Containment.
refueling outage.
          During a refueling outage, the reactor vessel head and steam
c.
          dryer /separatorr assemblies were removed from the reactor vessel without
Conclusions
          secondary containment integrity (SCI) established. This issue was
The control of a short duration mid-cycle outage was excellent.
          reviewed by the NRC Office of Nuclear Reactor Regulation. It was
08
          determined that the removal of the nead and assemblies without SCI
Miscellaneous Operations Issues (92700, 92901)
          established were not activities prohibited by TS 3.6.5.1. The potential
08.1 (Closed) Unresolved Item (URI) 50-325/96-15-01:
!         for load handling accidents was a safety cuestion that has been reviewed
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
'
'
          by the NRC. However, maintenance of SCI curing vessel disassably was a
logical extension of the defense-in-depth ap3 roach used in addressing
          logical extension of the defense-in-depth ap3 roach used in addressing
the heavy loads issue and encouraged by the
          the heavy loads issue and encouraged by the 4RC. The licensee's action
4RC.
          in proceeding with vessel disassembly was not conservative. The
The licensee's action
          licensee implemented controls during the Unit 2 refueling outage to
in proceeding with vessel disassembly was not conservative.
          maintain secondary containment operable during vessel disassembly. This
The
          issue was thoroughly evaluated as part of the licensee's Safe Shutdown
licensee implemented controls during the Unit 2 refueling outage to
          Risk Management Assessment.
maintain secondary containment operable during vessel disassembly. This
    08.2 (Closed) Violation V10 50-325(324)/97-02-01:                       Locked Valve Out of
issue was thoroughly evaluated as part of the licensee's Safe Shutdown
          Position
Risk Management Assessment.
          The licensee's response to this violation was dated May 5, 1997, and was
08.2 (Closed) Violation V10 50-325(324)/97-02-01:
          accepted by the NRC in a letter dated May 23. 1997. The corrective
Locked Valve Out of
          actions described in the response letter were verified as complete by
Position
          the inspector. This violation is closed.
The licensee's response to this violation was dated May 5, 1997, and was
    08.3 (Closed) URI 50-325/97-12-03:                   Recirculation Pumo Run backs
accepted by the NRC in a letter dated May 23. 1997.
          On November 5. 1997, the licensee began a c0ntrolled shutdown for the
The corrective
          Unit 1 forced outage in order to replace leaning fuel bundles. During
actions described in the response letter were verified as complete by
          the shutdown. Unit I received two recirculation pump run backs to the 45
the inspector.
          percent limiter. During the second run back the five percent buffer
This violation is closed.
          region was entered and exited in accordance with procedures.
08.3 (Closed) URI 50-325/97-12-03:
)         Subsequently. no other transients or run backs were ercountered while
Recirculation Pumo Run backs
          removing the Reactor Feedwater Pumps (RFPs) from service. The licensee
On November 5. 1997, the licensee began a c0ntrolled shutdown for the
          preliminarily attributed the first run back to a malfunction of the 1B
Unit 1 forced outage in order to replace leaning fuel bundles. During
          discharge check valve causing diversion of the 1A RFP through the 1B
the shutdown. Unit I received two recirculation pump run backs to the 45
          discharge valve to the main condenser. The final analysis was provided
percent limiter.
          in the root cause analysis for CR 97-3917. Unit 1 Plant Transients While
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
8
        removal of the RFPs at too high of a power level and a design problem in
Removing a Reactor Feed Pump from Service. The inspector reviewed the
        the a) plication of the Metal-On-Silicon Field Effect Transistor (MOSFET)
analysis and noted that the root cause attributed the run backs to the
        switcl. The MOSFET was used in the 45 percent recirculation pump run
removal of the RFPs at too high of a power level and a design problem in
        back logic to indicate the below 182 inches reacter water level contact
the a) plication of the Metal-On-Silicon Field Effect Transistor (MOSFET)
        which is one of two contacts required to initiate the run back.
switcl.
        Reactor water level perturbations are expected during the removal of the
The MOSFET was used in the 45 percent recirculation pump run
        RFPs from service: however the magnitude of these perturbations seen for
back logic to indicate the below 182 inches reacter water level contact
        these events were outside of the operators expectations. The root cause
which is one of two contacts required to initiate the run back.
        analysis stated that removal of the RFP at 65 percent power was
Reactor water level perturbations are expected during the removal of the
        inappropriate in that 65 percent during this evolution has changed since
RFPs from service: however the magnitude of these perturbations seen for
        power uprate. Before power uprate. RFPs were removed from service 3er
these events were outside of the operators expectations.
        10P-32, Condensate and Feedwater System Operating Procedure, at or )elow
The root cause
        65 percent. Under current conditions 65 percent is approximately
analysis stated that removal of the RFP at 65 percent power was
        equivalent to 68 percent power pre-uprated power. The analysis
inappropriate in that 65 percent during this evolution has changed since
        attributed the magnitude of the perturbations to removal at too high of
power uprate.
        a power level. In addition, the licensee determined that when the first
Before power uprate. RFPs were removed from service
        RFP was taken out of service, the less than 20 percent RFP flow contact
3er
        for the 18 pump was made up and with the MOSFET improperly indicating
10P-32, Condensate and Feedwater System Operating Procedure, at or )elow
        below 182 inches water level the run backs were received. The design of
65 percent.
        the MOSFET causes the contact to not be able to properly position itself
Under current conditions 65 percent is approximately
        u'aon loss of the constant voltage supply. Therefore interruptions in
equivalent to 68 percent power pre-uprated power.
        tle voltage will cause the MOSFET contact to not function as designed.
The analysis
        The second Run back was also attributed to the MOSFET.                     The licensee
attributed the magnitude of the perturbations to removal at too high of
        intends to replace the MOSFETs in the next Unit 1 outage, The inspector
a power level.
        noted that the MOSFETs had already been replaced in Unit 2.
In addition, the licensee determined that when the first
        The licensee is reviewing plant operation to determine the appropriate
RFP was taken out of service, the less than 20 percent RFP flow contact
        power level for removal of the RFPs from service. Based on licensee
for the 18 pump was made up and with the MOSFET improperly indicating
        satisfactory comaletion of the investigation into the cause for the
below 182 inches water level the run backs were received. The design of
        multiple run bac(s on November 5-6, 1997 this item is closed.
the MOSFET causes the contact to not be able to properly position itself
  08.4 (Closed) URI 50-325(324)/97-12-04:                     Diesel Doeration Low Voltace Auto
u'aon loss of the constant voltage supply.
        Start Defeated
Therefore interruptions in
        The inspector reviewed the licensee's root cause investigation CR 97-
tle voltage will cause the MOSFET contact to not function as designed.
        03683, 4KV Bus 2C/2D Clearances. The licensee's investigation
The second Run back was also attributed to the MOSFET.
        determined that the number 3 diesel generator (DG) undervoltage relay
The licensee
        had been disabled in the same manrer as the number 4 DG during similar
intends to replace the MOSFETs in the next Unit 1 outage, The inspector
        maintenance activities on different days.
noted that the MOSFETs had already been replaced in Unit 2.
          The inspector verified that the licensee did not exceed TS action,
The licensee is reviewing plant operation to determine the appropriate
          limiting condition for operation, or time requirements for both
power level for removal of the RFPs from service.
          electrical bus maintenance activities. The inspector found that, on
Based on licensee
          October 9. 1997, the plant was under a TS action statement requirement
satisfactory comaletion of the investigation into the cause for the
          per TS 3.8.2.1. to restore the inoperable bus to operable within 8
multiple run bac(s on November 5-6, 1997 this item is closed.
          hours, or be in hot shutdown within the next 12 hours. The electrical
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
bus was not restored, in this case, for 12 hours and 58 minutes.
          plant condition was not recognized as a problem until the root cause
This
          investigation was performed. The root cause investigation was found to
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
9
          recognize that the DG would be inoperable during the implementation of
be adequate.
          their clearance. This item is closed.
The ins)ector concluded that the licensee *s control of the
                                          II. Maintenance
2C and 2D electrical aus maintenance was weak because they did not
    M1   Conduct of Maintenance
recognize that the DG would be inoperable during the implementation of
    M1.1 Spent Fuel Cask Movement
their clearance.
      a. Inspection Scooe (62707)
This item is closed.
          The inspector observed transfer of the spent fuel shipaing cask from the.
II. Maintenance
          117 foot elevation to the transport v'hicle and from t1e transfer
M1
          vehicle to the 117 foot elevation of the Unit 1 Reactor Building.             s
Conduct of Maintenance
      b. Observations and Findinas
M1.1 Spent Fuel Cask Movement
          On December 8. 1997, the inspector observed the removal of the spent
a.
          fuel shipping cask, with fuel in the cask from the 117 foot to the 20
Inspection Scooe (62707)
          foot elevation in the Unit 1 Reactor Building. On December 15, 1997,
The inspector observed transfer of the spent fuel shipaing cask from the.
          the inspector observed shipping cask movement, without fuel in the cask,
117 foot elevation to the transport v'hicle and from t1e transfer
          from the 20 foot elevation to the 117 foot elevation in the Unit 1
vehicle to the 117 foot elevation of the Unit 1 Reactor Building.
          Reactor Building. During both evolutions the cask was transferred with
s
          the valve box covers removed while being moved by the non-single failure
b.
          proof yoke. Approval for use of a non-single failure proof yoke for
Observations and Findinas
          movement of the cask with the valve covers removed was granted to the-
On December 8. 1997, the inspector observed the removal of the spent
l         licensee by the NRC in a letter dated December 2, 1997. Upon reaching
fuel shipping cask, with fuel in the cask from the 117 foot to the 20
          the transfer vehicle on December 8. 1997. the cask was wiped down to
foot elevation in the Unit 1 Reactor Building.
          reduce contami.1ation levels. During both movements the inspector noted
On December 15, 1997,
          that the area was adequately posted for the radiological conditions
the inspector observed shipping cask movement, without fuel in the cask,
I         present and i ealth pnysics personnel were present. The inspector noted
from the 20 foot elevation to the 117 foot elevation in the Unit 1
          that adequate maintenance supervisory oversight was present for both
Reactor Building.
          cask movements.
During both evolutions the cask was transferred with
          Subsequent surveys of the cask after removal from the Reactor Building
the valve box covers removed while being moved by the non-single failure
          revealed that the shipment exceeded required limits. This event was
proof yoke. Approval for use of a non-single failure proof yoke for
          captured in CR 97-4161. S)ent Fuel Cask (IF-300). The cask was returned
movement of the cask with the valve covers removed was granted to the-
          to the Reactor Building w1ere additional decontamination was conducted.
l
          The licensee attributed the contamination levels seen to leaching of the
licensee by the NRC in a letter dated December 2, 1997. Upon reaching
          contamination due to changing temperatures and weather conditions.
the transfer vehicle on December 8. 1997. the cask was wiped down to
      c. Conclusions
reduce contami.1ation levels.
          Movement of the spent fuel shi) ping cask was performed in accordance
During both movements the inspector noted
          with methodology approved by t1e NRC in a letter dated December 2. 1997.
that the area was adequately posted for the radiological conditions
          Adequate supervisory oversight was present during movement of the cask.
I
                                                                                    -_ _a
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
10
    M1.2 RCIC Turbine Exhaust Diaphraam High Pressure Instrument Calibration
M1.2 RCIC Turbine Exhaust Diaphraam High Pressure Instrument Calibration
      a. Insoection Scoce (61726)
a.
          The inspector observed the performance of Maintenance Surveillance Test
Insoection Scoce (61726)
          2MST-RCIC230. RCIC Turbine Fxhaust Giaphragm High Pressure Instrument
The inspector observed the performance of Maintenance Surveillance Test
          Channel Calibration, for the pressure switches 2-E51-PSH-N012A and 2-
2MST-RCIC230. RCIC Turbine Fxhaust Giaphragm High Pressure Instrument
          E51-PSH-N012C.
Channel Calibration, for the pressure switches 2-E51-PSH-N012A and 2-
      b. Observations and Findinas
E51-PSH-N012C.
          On December 24. 1997, with Unit 2 at 100 percent power the inspector
b.
          observed the channel calibration for RCIC pressure switches 2 E51-PSH-
Observations and Findinas
          N012A and 2-E51-PSH-N012C.     The inspector verified that duriug the
On December 24. 1997, with Unit 2 at 100 percent power the inspector
          performance of this channel calibration that HPCI and Automatic
observed the channel calibration for RCIC pressure switches 2 E51-PSH-
          Depressurization System (ADS) were o)erable and that no othar work
N012A and 2-E51-PSH-N012C.
          activities were being conducted whic1 could cause an inadvertent
The inspector verified that duriug the
          isolation. This test verified that, upon sensing of a high pressure
performance of this channel calibration that HPCI and Automatic
          condition between the t'arbine exhaust dia)hragms, an isolation signal is
Depressurization System (ADS) were o)erable and that no othar work
          sent in accordance with TS 4.3.2.1 and Ta)les 3.3.2-2(4.b.6) and 4.3.2-
activities were being conducted whic1 could cause an inadvertent
          1(4.b.6)
isolation.
          The inspector reviewed the work request / job order (WR/J0) AKNU 19 and
This test verified that, upon sensing of a high pressure
          the governing procedure 2MST-RCIC230. The procedure in use was verified
condition between the t'arbine exhaust dia)hragms, an isolation signal is
          to be the correct revision and the test instrumentation in use was
sent in accordance with TS 4.3.2.1 and Ta)les 3.3.2-2(4.b.6) and 4.3.2-
          within the allowable calibration duration. The inspector observed the
1(4.b.6)
'        procedure in use at all work locations and adequate communication was
The inspector reviewed the work request / job order (WR/J0) AKNU 19 and
          maintained throughout the test. The work observed was completed
the governing procedure 2MST-RCIC230. The procedure in use was verified
          satisfactorily with no observed concerns.
to be the correct revision and the test instrumentation in use was
    c.   Conclusions
within the allowable calibration duration.
          The inspector observed performance of cal:uration of two RCIC pressure
The inspector observed the
          switches. The work activities were completed without any identified
l
          questions or concerns.
procedure in use at all work locations and adequate communication was
    M1.3 General Comments
'
      a. Insoection Scone (62700)
maintained throughout the test.
          The inspector examined the following work activities involving EQ
The work observed was completed
          electrical equipment to verify maintenance implementation of EQ
satisfactorily with no observed concerns.
          requirements.
c.
          *
Conclusions
                  WR/JO 97-ALVT-002 Verified Calibration of Unit 1 Loop B Residual
The inspector observed performance of cal:uration of two RCIC pressure
                  Heat Removal (RHR) Service Water Pressure Switches Tag No.
switches.
                  1-SW-PS-1176 B and 1-SW-PS-11760
The work activities were completed without any identified
          *
questions or concerns.
                  WR/JO 97-AGDR-002 Verified Calibration of Unit 1. Loop A. RHR Flow
M1.3 General Comments
                  Transmitter (1-E11-FT-N015A). Converter (1-E11-FY-5119A). Square
a.
                  Root Converter (1-E11-FY-K600A)
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
11
        .
WR/JO 97-AAAS-002 Unit 2. Loop B. RHR Breaker Test in compartment
                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
DM 5 of GE IC 7700 Series MCC 2XB-2 Division Il
    b. Observations and findinos
b.
        The above work was ,m cformed with the work packages present and in
Observations and findinos
        active use. Technicians were skillful, experienced, and knowledgeable
The above work was ,m cformed with the work packages present and in
        of their assigned tasks. However, on December 10, 1997, while observing
active use.
        Instrumentation and Control (I&C) maintenance personnel perform work
Technicians were skillful, experienced, and knowledgeable
        activities in accordance with WR/JO 97-AAAS-002, the inspector noted
of their assigned tasks.
        that one of the multiple cable electrical penetrations in the top of MCC
However, on December 10, 1997, while observing
        2-2XB-2 did not have Nelson flame guard putty on the inside surface as
Instrumentation and Control (I&C) maintenance personnel perform work
        required by Maintenance Procedure OMMM 016. Environmental Qualification
activities in accordance with WR/JO 97-AAAS-002, the inspector noted
        Maintenance Program. Revision 4. to properly seal the penetration. The
that one of the multiple cable electrical penetrations in the top of MCC
        inspector examined the putty installation on the top of the MCC cabinet
2-2XB-2 did not have Nelson flame guard putty on the inside surface as
        for each of the penetrations and found the putty seal severely damaged
required by Maintenance Procedure OMMM 016. Environmental Qualification
        on a second multiple cable penetration. In addition, cables were loose
Maintenance Program. Revision 4. to properly seal the penetration.
        in both of the multiple cable penetrations. The applicable
The
        Environmental Otalification Data Package (ODP). ODP 67, requires missing
inspector examined the putty installation on the top of the MCC cabinet
        or disturbed Nelson putty seals to be repaired or replaced. However,
for each of the penetrations and found the putty seal severely damaged
        the PM procedure used to maintain and inspect the MCC's (PM Procedure
on a second multiple cable penetration.
        OPM-MCC002. Revision 7. PM of GE Motor Control Centers and Switchboards)
In addition, cables were loose
        did not have inspection requirements or acceptance criteria to ensure
in both of the multiple cable penetrations.
        that putty seals were properiy sealing the cabinets. On September 17.
The applicable
f       1997, a three-year PM conducted on MCC 2-2XB-2 would have identified
Environmental Otalification Data Package (ODP). ODP 67, requires missing
l       this discrepancy had procedure OPM-MCC002 included the acceptance
or disturbed Nelson putty seals to be repaired or replaced.
        criteria for the Nelson flame seal putty. A subsequent inspection
However,
        performed on December 11. 1997 by the licensee, of 22 MCCs found an
the PM procedure used to maintain and inspect the MCC's (PM Procedure
        additional three MCC cabinet penetrations with damaged Nelson putty
OPM-MCC002. Revision 7. PM of GE Motor Control Centers and Switchboards)
        seals. In addition. 15 3ercent of the cables inspected in cabinet
did not have inspection requirements or acceptance criteria to ensure
        penetrations had putty w1ich appeared not to fully adhere to the cable
that putty seals were properiy sealing the cabinets.
        in some areas. Failure of the procedure to implement E0 requirements
On September 17.
        for Nelson autty seals is identified as VIO 50-325(324)/97-13-02.
f
        Inadequate 3rocedure for the Conduct of E0 Preventive Maintenance.
1997, a three-year PM conducted on MCC 2-2XB-2 would have identified
    c. Conclusions
l
        Maintenance activities observed related to E0 of electrical equiament
this discrepancy had procedure OPM-MCC002 included the acceptance
        were found to be conducted in a thorough and effective manner, iowever,
criteria for the Nelson flame seal putty. A subsequent inspection
        a violation was identified for a PM procedure not indicating specific E0
performed on December 11. 1997 by the licensee, of 22 MCCs found an
        requirements. This omission resulted in deficient Nelson flame seals in
additional three MCC cabinet penetrations with damaged Nelson putty
        MCCs not being dettcted during scheduled PM activities.
seals.
  M3   Maintenance Procedures and Documentation
In addition. 15 3ercent of the cables inspected in cabinet
  M3.1 Steam Jet Air Eiector Off-Gas Radiation Monitor increase
penetrations had putty w1ich appeared not to fully adhere to the cable
    a. Inspection Scoce (61726)
in some areas.
        The inspector reviewed selected sections of Operating Instruction 101-
Failure of the procedure to implement E0 requirements
        03.1. Control Operator Daily Surveillance Report to ensure that
for Nelson autty seals is identified as VIO 50-325(324)/97-13-02.
                                                                                  i
Inadequate 3rocedure for the Conduct of E0 Preventive Maintenance.
                                                                                  1
c.
                                                                                  ,
Conclusions
            e
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
12
      appropriate and prompt actions were taken to address abnormal TS
appropriate and prompt actions were taken to address abnormal TS
      surveillance values,
surveillance values,
    b. Observations and Findinos
b.
      On December 2. 1997. Unit 1 was in mode 1 at 100 percent power.         The
Observations and Findinos
      inspector reviewed the daily surveillance report as contained in
On December 2. 1997. Unit 1 was in mode 1 at 100 percent power.
      Attachment 1 to 101-03.1 for November 30 through December 1. 1997.           The
The
      inspector noted that the values for the Steam Jet Air Ejector (SJAE)
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)
off-9as radiation monitors on aage 26 were between 1570 and 1780
i
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
      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
value of 1000 mR/hr.
      the detection of fuel element failures. The radiation levels are
The SJAE off-gas radiation monitors provide for
      recorded in 101-03.1 to provide an indication whether SJAE off-gas
the detection of fuel element failures.
      radiation levels are approaching the alarm setpoint, which serves to
The radiation levels are
      ensure that dose rates for gaseous effluents do not exceed the limits
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
l
      prescribed in TS 3.11.2.1. Dose Rate.
The inspector reviewed the associated procedures, work tickets, and
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
      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
      required the control operator to red circle all values wt           are not
are not
      within required limits. The inspector noted no indication on the
within required limits.
      attachment or in the operator logs that action had been taken or was
The inspector noted no indication on the
      expected to be performed to address the out-of-range values. Subsequent
attachment or in the operator logs that action had been taken or was
      reviews of the daily log entries by the inspector indicated continual
expected to be performed to address the out-of-range values.
      abnormal values and no red circles.         These failures were recorded in CR
Subsequent
      97-4136. Daily Surveillance Report.         The failure to red circle values
reviews of the daily log entries by the inspector indicated continual
      not within required limits is a violation.         This violation is identified
abnormal values and no red circles.
      as VIO 50-325/97 13-03. Failure to Note Abnormal TS Surveillance Values.
These failures were recorded in CR
      CR 97-4100. Questioned OG Data / Fuel Leak       indicated that on December 3,
97-4136. Daily Surveillance Report.
      1997, a step increase of approximately 200 mR/hr was seen on the
The failure to red circle values
      radiation monitor Subsequent sample results have shown an increase in
not within required limits is a violation.
      the Sum of Six value ano changes in the fuel reliability index which are
This violation is identified
      signs of potential fuel failure. In addition, the inspector noted that
as VIO 50-325/97 13-03. Failure to Note Abnormal TS Surveillance Values.
        incorrect sensitivities were used during the November 25, 1997.
CR 97-4100. Questioned OG Data / Fuel Leak
      adjustment of the SJAE radiation monitor alarm setpoilts. This was
indicated that on December 3,
      documented by the licensee in CR 97-4046. SJAE Rad Mci. sensitivities.
1997, a step increase of approximately 200 mR/hr was seen on the
      CR 97-4180 SJAE rad monitor setpoints, addressed coordination problems           ',
radiation monitor
      between the Operations procedure used to request new radiation monitor
Subsequent sample results have shown an increase in
      setpoints, the Environmental and Radiological Control (E&RC) proced ce
the Sum of Six value ano changes in the fuel reliability index which are
      that calculates the new setpoint, and the Maintenance procedure that
signs of potential fuel failure.
        installs the new setpoints. By the time the radiation monitor setpoints
In addition, the inspector noted that
      were ready to be installed the new values needed to be recalculated.
incorrect sensitivities were used during the November 25, 1997.
      The inspector determined as a result of the cited failure and the three
adjustment of the SJAE radiation monitor alarm setpoilts.
      additional CRs mentioned previously, that control and monitor'.ng of the
This was
      alarm setpoint was poor. Previous instances of failing to properly
documented by the licensee in CR 97-4046. SJAE Rad Mci. sensitivities.
      disposition abnormal values were recorded by the NRC in Inspection
CR 97-4180 SJAE rad monitor setpoints, addressed coordination problems
      Re) ort (IR) 50-325(324)/97-12, when inadequate corrective action was
between the Operations procedure used to request new radiation monitor
      tacen for abnormally high drywell temperature. Tne abnormal temperature
',
        resulted in exceeding the calculated environmental limits for ten
setpoints, the Environmental and Radiological Control (E&RC) proced ce
        snubbers in the drywell.
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
13
        c. Conclusions
c.
            The licensee continues to struggle with proper dispositioning of
Conclusions
            abnormal indications. The failure to maintain the Daily Surveillance
The licensee continues to struggle with proper dispositioning of
            Report in accordance with procedure was a violation. Abnormal values
abnormal indications.
            observed for the Steam Jet Air Ejector radiation monitor and subsequent
The failure to maintain the Daily Surveillance
            test indicate potential fuel failure for Unit 1.
Report in accordance with procedure was a violation. Abnormal values
      M8   Miscellaneous Maintenance Issues (92902)
observed for the Steam Jet Air Ejector radiation monitor and subsequent
      M8.1 (Closed) Licensee Event Reoort (LER) 50-325(324)/96-017-00:                                   Invalid
test indicate potential fuel failure for Unit 1.
            Loss of Coolant Accident Locic Actuation
M8
            The invalid LOCA. initiation signal occurred during installation of test
Miscellaneous Maintenance Issues (92902)
            equipment to support surveillance testing.                                 P16nt systems responded as
M8.1
            designed. The initiation signal resulted in the following actuation:
(Closed) Licensee Event Reoort (LER) 50-325(324)/96-017-00:
                  Automatic start of emergency DGs 1.2.3. and 4.
Invalid
                  Automatic start of Unit 1 Core Spray (CS) pump 1A.
Loss of Coolant Accident Locic Actuation
                  Automatic start of Unit 2 Nuclear Service Water (NSW) pump 2A.
The invalid LOCA. initiation signal occurred during installation of test
                  Unit 1 Grou) 10 division 1 actuation.
equipment to support surveillance testing.
                  Closure of Jnit 1 Reactor Building Closed Cooling Water heat
P16nt systems responded as
                  exchanger Service Water isolation valve.1-SW-V106.
designed. The initiation signal resulted in the following actuation:
                  0)ening of NSW header to vital header isolation valve. 1-SW-V117.
Automatic start of emergency DGs 1.2.3. and 4.
,                  Slutdown of 1A and 10 Unit 1 drywell coolers                                                       ;
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
1
            Corrective actions, described in the LER. were reviewed and verified by
Corrective actions, described in the LER. were reviewed and verified by
            the inspector. -These included: appropriate administrative action with
the inspector. -These included: appropriate administrative action with
            the involved technician; briefing of maintenance 1&C technicians on this
the involved technician; briefing of maintenance 1&C technicians on this
            event; providing maintenance I&C personnel managements expectations ft
event; providing maintenance I&C personnel managements expectations ft
            the restart of surveillance tests after problems have been encountered;
the restart of surveillance tests after problems have been encountered;
            restricting the use of Simpson Model 260 Voltage Ohm Meters (V0Ms) for
restricting the use of Simpson Model 260 Voltage Ohm Meters (V0Ms) for
            circuit checks specified in maintenance surveillance tests: developing
circuit checks specified in maintenance surveillance tests: developing
            training to enhance technician knowledge of the effects of test
training to enhance technician knowledge of the effects of test
            equipment misalignment: and revising maintenance procedures to preclude
equipment misalignment: and revising maintenance procedures to preclude
            similar events.
similar events.
            This event did not violate TS. This LER is closed.
This event did not violate TS. This LER is closed.
      M8.2 (Closed) LER 50-325/97-009-00: Missed Increased Frecuency Inservice
M8.2 (Closed) LER 50-325/97-009-00: Missed Increased Frecuency Inservice
            Testino Recuirement
Testino Recuirement
            The American Society of Mechanical Engineers (ASME) Boiler and Pressure
The American Society of Mechanical Engineers (ASME) Boiler and Pressure
            Vessel Code. Section XI, 1980 Edition through Winter 1981. Addenda
Vessel Code. Section XI, 1980 Edition through Winter 1981. Addenda
            Section IWV-3414(a), requires an increase in test frequency in the event
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
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
is observed.
            increased as required. The required testing was missed by about two
Contrary to this requirement, the test frequency was not
            weeks. Upon discovery. the valve was tested and the stroke time was
increased as required.
            within the previous value and the test met the ASME Section XI
The required testing was missed by about two
            requirements.
weeks.
                                                                                                                        !
Upon discovery. the valve was tested and the stroke time was
                                                                                                                        l
within the previous value and the test met the ASME Section XI
                                                                                                                  ___J
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       ;
i
              testing of the RBM instrument channel 3 int                                 '
14
                                                                                              : tion, This condition
s
              has existed since November 1996 for Unit 1, ma December 1996 for
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
'
'
              Unit 2. Upon discovery, the correct tests were performed on both units
which indicated that the equipment was in calibration and capable of
              which indicated that the equipment was in calibration and capable of
performing its safety function.
              performing its safety function.
The error was attributed to an inadequate administrative review of
              The error was attributed to an inadequate administrative review of                                       ;
;
              reformatting changes made in September 1996. The surveillance procedure
reformatting changes made in September 1996. The surveillance procedure
              changes were being upgraded in accordance with the generic procedure
changes were being upgraded in accordance with the generic procedure
              writers guide. However, these changes did not insert the proper steps
writers guide.
              to test the RBM inop instrument channel B.
However, these changes did not insert the proper steps
            ' Corrective actions, described in the LER. were reviewed and verified by
to test the RBM inop instrument channel B.
            -the inspector. The inspector determined that this event did not violate
' Corrective actions, described in the LER. were reviewed and verified by
              TS since only the test for channel B was missed. The situation was
-the inspector.
              corrected within the allowable time specified by TS 3/4.3.4.
The inspector determined that this event did not violate
              The-results of the RBM inop functional tests performed on toth units
TS since only the test for channel B was missed. The situation was
              upon discovery, indicated that the equipment was in calibration and
corrected within the allowable time specified by TS 3/4.3.4.
              capable of performing its intended safety function. This LER is closed.
The-results of the RBM inop functional tests performed on toth units
        M8.4 (Closed) LER 50-325(324)/95-022-00:                                   HPCI System Discharae Flow Element
upon discovery, indicated that the equipment was in calibration and
              Gasket Leak
capable of performing its intended safety function.
                During performance of a post maintenance test on the HPCI system. the
This LER is closed.
              discharge flow element flanged gasket developed a 5 to 10 gallons per
M8.4 (Closed) LER 50-325(324)/95-022-00:
                minute (gpm) leak. Several other problems were also observed with
HPCI System Discharae Flow Element
                system operation.
Gasket Leak
                Investigation revealed that undersized flange studs had been originally
During performance of a post maintenance test on the HPCI system. the
                installed on the flow element flange, allowing the Flexitallic gasket to
discharge flow element flanged gasket developed a 5 to 10 gallons per
                be installed off center. The off centered gasket degraded during the
minute (gpm) leak. Several other problems were also observed with
                post maintenance test. This condition existed on both units and
system operation.
                prompted declaring a potential failure of the HPCI system to ]erform its
Investigation revealed that undersized flange studs had been originally
                intended safety function.                               With the HPCI system inoperable tie TS
installed on the flow element flange, allowing the Flexitallic gasket to
                                                                                                                      U
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
15
            oermitted continued reactor operation provide 1 the ADS. CS system, and
oermitted continued reactor operation provide 1 the ADS. CS system, and
            RCIC were operable. This event was withir, Me TS requirement.
RCIC were operable. This event was withir, Me TS requirement.
            Corrective measures as described in the LER were reviewed and verified
Corrective measures as described in the LER were reviewed and verified
            by the inspector. This LER is closed.
by the inspector.
      M8.5 (Closed) Ins)ection Follow-un item (IFI) 50-325/97-05-02:                 Abnormal CS
This LER is closed.
            Soarcer Brea t Detector Indication
M8.5 (Closed) Ins)ection Follow-un item (IFI) 50-325/97-05-02:
            (Closed) VIC 50-325/97-06-03:               Inadeauate CS Surveillance Procedure
Abnormal CS
            .(Closed) LER 50-325/97 02:               Core Soray Header Differential Pressure
Soarcer Brea t Detector Indication
            Instrumentation InoDerable
(Closed) VIC 50-325/97-06-03:
            On March 9.1997, en auxiliary o)erator (AO) was verifying
Inadeauate CS Surveillance Procedure
            instrumentation indications in tie Unit 1 Reactor Building.                 The A0
.(Closed) LER 50-325/97 02:
            observed.that the reading displayed for 1-E21-PDS-N004A. Core Spray Line
Core Soray Header Differential Pressure
            Break Indicator, was not within TS 4.5.3.1.2.c.2 requirements. This
Instrumentation InoDerable
              )ressure switch functioned to detect a break in the CS piping located
On March 9.1997, en auxiliary o)erator (AO) was verifying
l           3etween the vessel and the shroud. The differential pressure (dP)
instrumentation indications in tie Unit 1 Reactor Building.
            sensor measures the pressure across the core. Due to the addition of
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)
'
'
L            the drop from the steam separator, any break in the line would cause the
sensor measures the pressure across the core.
l    -
Due to the addition of
            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
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
Themfore. the licensee could not evaluate whether the alarm setpoint
I
I
            Themfore. the licensee could not evaluate whether the alarm setpoint
was within the " normal" TS range.
            was within the " normal" TS range. This nonconformance resulted in VIO
This nonconformance resulted in VIO
            50 325/97-05-02. Inadequate CS Surveillance Procedure.
50 325/97-05-02. Inadequate CS Surveillance Procedure.
            The licensee performed reviews of data collected nonroutinely during
The licensee performed reviews of data collected nonroutinely during
              1995-1996 and in ESR 97-181 calculated a " normal" value for setpoint
1995-1996 and in ESR 97-181 calculated a " normal" value for setpoint
            verification in the related surveillance procedures. The licensee
verification in the related surveillance procedures.
            subsecuently changed the alarm setpoints and updated the affected
The licensee
            procec ures. Additionally. the licensee performed a review of the TS and
subsecuently changed the alarm setpoints and updated the affected
            determined that appropriate logging of required TS values was being
procec ures. Additionally. the licensee performed a review of the TS and
            accomplished. During the refueling outage for Unit 2 from Se]tember to
determined that appropriate logging of required TS values was being
            October 1997 the licensee, with prior NRC approval, uprated t1e 100
accomplished. During the refueling outage for Unit 2 from Se]tember to
            percent _ rated thermal power 5 percent. The licensee included
October 1997 the licensee, with prior NRC approval, uprated t1e 100
              verification of CS sparger dP " normal" values as part of the uprate
percent _ rated thermal power 5 percent. The licensee included
              test program performed in accordance with S)ecial Procedure 2SP-97-204.
verification of CS sparger dP " normal" values as part of the uprate
              Unit 2 Power Jprate Data Collection. The cleck served to record the CS
test program performed in accordance with S)ecial Procedure 2SP-97-204.
              sparger shutdown values.
Unit 2 Power Jprate Data Collection. The cleck served to record the CS
              The inspector reviewed ESR 97-634. ESP-97-204. CR 97-3870. LER 50-
sparger shutdown values.
              325/97-02, and other related documentation. The inspector verified that
The inspector reviewed ESR 97-634. ESP-97-204. CR 97-3870. LER 50-
              routine recording. upon entering mode 1. of the CS sparger dP was
325/97-02, and other related documentation.
              incorporated into 0)erating Instruction 001-03.3. Auxiliary Operator
The inspector verified that
              Daily Surveillance Report for both units. CR 97-3870. Core Spray Leak
routine recording. upon entering mode 1. of the CS sparger dP was
              Detection, documented the discovery on October 29, 1997 by an AD, that
incorporated into 0)erating Instruction 001-03.3. Auxiliary Operator
                                                                                                  4
Daily Surveillance Report for both units.
                                                                                                  l
CR 97-3870. Core Spray Leak
Detection, documented the discovery on October 29, 1997 by an AD, that
4
l


                            _ - _ _ _ _ _   - - _ - _ _ _ _ _ _ _
_ - _ _ _ _ _
- - _ - _ _ _ _ _ _ _
'
'
                                                                  16
16
        the 2-E21-PDS-N004A. CS A Loop Leak Detection, was outside of its
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
specified range.
        entered. The licensee determined the new CS dP range in ESR 97-634
The instrument was declared inoperable and an LC0 was
        Core Spr 3y Loop Line Breuk Detectio , Allowable Range Change. The new
entered.
        alarm setpoints were implemented and integrated into the affected
The licensee determined the new CS dP range in ESR 97-634
        surveillances. 3rocedures, and design documents. Based on completion of
Core Spr 3y Loop Line Breuk Detectio , Allowable Range Change.
        the review of t1e TS for other " normal" values not properly trended,
The new
        adjustment of the dP alarm setpoints*to bring the setpoints into
alarm setpoints were implemented and integrated into the affected
        rvpliance with TS. and the institution of routine monitcring of the CS
surveillances.
        .qarger " normal" values these items are closed.
3rocedures, and design documents.
  M8.6 (Clos (d) VIO 50-325(324)/97-02-04: Failure to Imolement the
Based on completion of
        Renuirements of (a)(1) and (a)(2) of 10 CFR 50.65. The Maintenance Rule
the review of t1e TS for other " normal" values not properly trended,
        This violation reported that all historical data since July 10. 1993.
adjustment of the dP alarm setpoints*to bring the setpoints into
        had not been obtained to establish baseline system / structure / component
rvpliance with TS. and the institution of routine monitcring of the CS
        (SSC) performance, validate scoping, and set initial condition (a)(1)
.qarger " normal" values these items are closed.
        and condition (a)(2) in the case of the reactor protection system (RPS),
M8.6 (Clos (d) VIO 50-325(324)/97-02-04:
        Only corrective work. requests / job orders had been used for initial
Failure to Imolement the
        determination of functional failures. Therefore, instrument out-of-
Renuirements of (a)(1) and (a)(2) of 10 CFR 50.65. The Maintenance Rule
        calibration data had not been reviewed for the period of July 10. 1993
This violation reported that all historical data since July 10. 1993.
        through October 30. 1995. As an action related to Maintenance Rule
had not been obtained to establish baseline system / structure / component
        implementation. Procedure OMMM-004. PM. was revised on October 30. 1995,
(SSC) performance, validate scoping, and set initial condition (a)(1)
        to require that out-of-calibration data be evaluated for Maintenance
and condition (a)(2) in the case of the reactor protection system (RPS),
        Rule functional failure applicability. However, this requirement only
Only corrective work. requests / job orders had been used for initial
        collected subsequent instrument out-of-calibration data.
determination of functional failures.
        As corrective action for this violation, the licensee reviewed all
Therefore, instrument out-of-
        available instrument out-of-calibration data for the RPS and other
calibration data had not been reviewed for the period of July 10. 1993
        components / systems which support the Maintenance Rule functions.
through October 30. 1995. As an action related to Maintenance Rule
        Functional failures identified were evaluated against performance
implementation. Procedure OMMM-004. PM. was revised on October 30. 1995,
        criteria to determine whether (a)(1) status should be assigned.
to require that out-of-calibration data be evaluated for Maintenance
        Although six condition reports were issued to evaluate additional
Rule functional failure applicability.
        functional failures, no system was required to be classified (a)(1)
However, this requirement only
        based on this review. The inspector reviewed the licensee's corrective
collected subsequent instrument out-of-calibration data.
        actions and held discussions with a)plicable management and engineering
As corrective action for this violation, the licensee reviewed all
        personnel concerning this issue. T1e inspector concluded that the
available instrument out-of-calibration data for the RPS and other
        licensee had taken the necessary corrective action to correct the
components / systems which support the Maintenance Rule functions.
        deficient condition and had taken appropriate corrective action to
Functional failures identified were evaluated against performance
        prevent its recurrence.         This item is closed.
criteria to determine whether (a)(1) status should be assigned.
                                          III. Enaineerina
Although six condition reports were issued to evaluate additional
  El     Conduct of Engineering
functional failures, no system was required to be classified (a)(1)
  El.1 Review of Enaineerina Procedures
based on this review.
    a.   Insoection Scoce (37550)
The inspector reviewed the licensee's corrective
        The inspectors reviewed the licensee's procedures which control the
actions and held discussions with a)plicable management and engineering
          environmental qualification program.
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
  4
4
                                                        17
17
      b. Observations and Findinas
b.
        :The inspectors reviewed the procedures listed below which control
Observations and Findinas
          various activities related to the environmental qualification 3rogram to
:The inspectors reviewed the procedures listed below which control
          determine if the procedures implement the requirements of 10 C:R 50.
various activities related to the environmental qualification 3rogram to
          Appendix B. and 10 CFR 50.49. The following procedures were reviewed:
determine if the procedures implement the requirements of 10 C:R 50.
          EGR-NGGC 0005. Engineering Service Requests. Rev     6. dated
Appendix B. and 10 CFR 50.49.
          Septembe" 5. 1997
The following procedures were reviewed:
          EGR-NGGC-0007. Maintenance of Design Documents, Rev. 2. dated
EGR-NGGC 0005. Engineering Service Requests. Rev
          August 22, 1997
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
EGR-NGGC 0153. Engineering Instrument Setpoints. Rev. 3. dated
:         August 22. 1997
:
l         EGR-NGGC-0156. Environmental Qualification of mlectrical Equipment
August 22. 1997
l         Important to Safety. Rev. 4. dated October 8.1997
l
          ENP-13.6 Equipment Data Base System. Control and Revision
EGR-NGGC-0156. Environmental Qualification of mlectrical Equipment
          Rev. 12. dated June 25. 1997
l
          MCP-NGGC-401.. Material Acquisition (Procurement Receiving, and
Important to Safety. Rev. 4. dated October 8.1997
          Shipping). Rev. 3. dated August 26, 1997
ENP-13.6 Equipment Data Base System. Control and Revision
          The inspectors verified that the procedures provided adequate
Rev. 12. dated June 25. 1997
          instructions for establishing, maintaining and implementing the
MCP-NGGC-401.. Material Acquisition (Procurement Receiving, and
          requirements of'10 CFR 00.49 except for the issues discussed
Shipping). Rev. 3. dated August 26, 1997
          below.
The inspectors verified that the procedures provided adequate
          Section 9.6 of procedure EGR-NGGC-0156 provided the guidance for
instructions for establishing, maintaining and implementing the
          maintaining E0 qualification data packages (ODPs). The procedure
requirements of'10 CFR 00.49 except for the issues discussed
          specified that changes to ODPs are to be captured using the ESR
below.
          process. The procedure required that ODPs were to be periodically
Section 9.6 of procedure EGR-NGGC-0156 provided the guidance for
          updated as necessary to maintain auditability, to incorporate new
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
'
'
          requirements, to meet plant specific requirements, ard to keep the
number of outstanding-changes at a reasonable level.
          number of outstanding-changes at a reasonable level. However
However
5
5
          procedure EGR-NGGC-0156 did not specify a clear time requirement
procedure EGR-NGGC-0156 did not specify a clear time requirement
          for updating the CDPs. The inspectors also determined that
for updating the CDPs.
          procedure EGR-NGGC-0007 did not provide any requirements for
The inspectors also determined that
          updating ODPs.   The failure to s]ecify specific criteria in
procedure EGR-NGGC-0007 did not provide any requirements for
          procedures could result in the 0)Ps becoming unauditable which is
updating ODPs.
          contrary to the requirements of 10 CFR 50.49. The failure to
The failure to s]ecify specific criteria in
          maintain and u]date the ODPs was one of the causes of the
procedures could result in the 0)Ps becoming unauditable which is
          violation whic1 resulted in the civil penalty identified in NRC
contrary to the requirements of 10 CFR 50.49. The failure to
          Inspection Report (IR) 50-325(324)/96-14. The failure to
maintain and u]date the ODPs was one of the causes of the
          establish clear, definite requirements for updating ODPs was
violation whic1 resulted in the civil penalty identified in NRC
          identified as a violation example at the Shearon Harris Nuclear
Inspection Report (IR) 50-325(324)/96-14.
          Plant in NRC IR 50-400/97-12. Since all Brunswick 00Ps are being
The failure to
          revised and updated at the current time, a violation was not
establish clear, definite requirements for updating ODPs was
          identified for this issue during the current inspection. The
identified as a violation example at the Shearon Harris Nuclear
          licensee's corrective actions for the Harris plant will resolve
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
18
      plants use the same corporate EGR-NGGC ?,ocedures.
this problem since the Harris. Brunswick, n.d H. B. Robinson
      Procedure EGR NG D 0153 provides the methodology to establish
plants use the same corporate EGR-NGGC ?,ocedures.
      instrument setpoint margins sufficient to account for various
Procedure EGR NG D 0153 provides the methodology to establish
      instrument uncertainties and environmental effects including
instrument setpoint margins sufficient to account for various
      temperature, pressure, radiation, seismic, and insulation
instrument uncertainties and environmental effects including
      resistance errors
temperature, pressure, radiation, seismic, and insulation
      Although procedure EGR-NGGC-0153 provided guidance on the
resistance errors
      treatment of environmental effects, the inspectors noted that in
Although procedure EGR-NGGC-0153 provided guidance on the
      the discussion of temperature effects, the applicability of vendor 3
treatment of environmental effects, the inspectors noted that in
      worst case performance specifications to plant specific conditions i
the discussion of temperature effects, the applicability of vendor
      was not clear. The inspectors also noted that requirements for
3
      seismic effects in procedure EGR-NGGC-0153 were not clear
worst case performance specifications to plant specific conditions
      regarding t6 match / confirmation of vendor profiles to plant
i
      specific [     les or configuration,
was not clear.
      in addition, the inspectors noted that procedure EGR-NGGC-0153
The inspectors also noted that requirements for
      referenced Drawing 0-03056. Service Environment Chart Normal &
seismic effects in procedure EGR-NGGC-0153 were not clear
      Accident Conditions. Units 1 & 2. for information on accident
regarding t6
      temperature data to be used in instrument setpoint calculations.
match / confirmation of vendor profiles to plant
      The inspectors determined that-Drawing D-03056 was " frozen" on
specific [
      December 12. 1996, and was not available for use. The reason for
les or configuration,
      removal of Drawing 0-03056 from use was documented in CR 96-04002
in addition, the inspectors noted that procedure EGR-NGGC-0153
      which identif9d the need to revise. and update Drawing D-03056-to
referenced Drawing 0-03056. Service Environment Chart Normal &
      incorporate f icironmental data from the Reactor Building
Accident Conditions. Units 1 & 2. for information on accident
      Environmentai Renort (RBER), Revision 5. The inspectors noted in
temperature data to be used in instrument setpoint calculations.
      review of calculations initiated since December 1996, the RBER
The inspectors determined that-Drawing D-03056 was " frozen" on
      was referenced for temperature profiles in the re:ctor building.
December 12. 1996, and was not available for use. The reason for
      The licensee indicated that a revision to EGR-NGGC-0153 will be
removal of Drawing 0-03056 from use was documented in CR 96-04002
      initiated to resolve inconsistency in wording regarding the
which identif9d the need to revise. and update Drawing D-03056-to
      application of accident temperature / seismic effects to make it
incorporate f icironmental data from the Reactor Building
      clear that vendor test results would fully envelope site specific
Environmentai Renort (RBER), Revision 5.
      profiles unless an evaluation has been aerformed to evaluate the-
The inspectors noted in
      differences. Additional guidance will 3e included to characterize
review of calculations initiated since December 1996, the RBER
      the requirements for engineering reviews of test-data to ensure
was referenced for temperature profiles in the re:ctor building.
      seismic and environmental profiles are bounding for site specific
The licensee indicated that a revision to EGR-NGGC-0153 will be
      conditions. The licensee indicated procedure EGR-NGGC-0153 will
initiated to resolve inconsistency in wording regarding the
      also be revised to either remove D-03056 as the reference for
application of accident temperature / seismic effects to make it
      temperature data and replace it with the appropriate reference
clear that vendor test results would fully envelope site specific
      (the RBER) . or to correct the drawing.
profiles unless an evaluation has been aerformed to evaluate the-
      The inspectors also identified that procedure EGR-NGGC-0153 unde-
differences.
      Section 9.5.1. Calibration Errors, was not clear regarding
Additional guidance will 3e included to characterize
      instrument calibration surveillance requirements for as-left, as-
the requirements for engineering reviews of test-data to ensure
      found or leave-alone zone tolerances. The licensee indicated
seismic and environmental profiles are bounding for site specific
      that procedure EGR-NGGC-0153. Section 9.5.1. would be revised to
conditions. The licensee indicated procedure EGR-NGGC-0153 will
      clarify these requirements to indicate that calibration tolerances
also be revised to either remove D-03056 as the reference for
      are the defined limits, above and below a desired value, within
temperature data and replace it with the appropriate reference
      which an instrument loop signal may vary and not require
(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
a
                                                                            j
j


                      _..__-____- ___-_____-__ _____- -
_..__-____- ___-_____-__ _____- -
                                                                                                    .
.
                                                                                                            ..
..
                                                                                                                                ..
..
  4
4
.
.
                                                                      19
19
          adjustment.                                 Licensee engineers stated that calibration tolerances
adjustment.
          are understood to be "as-left" values.
Licensee engineers stated that calibration tolerances
          The inspectors will review Procedure EGR-NGGC-0153 in a future
are understood to be "as-left" values.
          inspection to followup on these issues. An ins)ector followup
The inspectors will review Procedure EGR-NGGC-0153 in a future
          item (IFI). 50-325(324)/97-13 06. Revisions to )rocedure EGR-NGGC-
inspection to followup on these issues. An ins)ector followup
          0153, was identified to the licensee pending further review by
item (IFI). 50-325(324)/97-13 06. Revisions to )rocedure EGR-NGGC-
          liRC.
0153, was identified to the licensee pending further review by
      c. Conclusions
liRC.
          With the exception of the issues discussed above, the inspectors
c.
          concluded that the licensee's procedures for implementation of the
Conclusions
          Environmental Qualification com) lied with the requirements of 10 CFR
With the exception of the issues discussed above, the inspectors
          50.49 and 10 CFR 50. Appendix 3. An IFI was identified to review
concluded that the licensee's procedures for implementation of the
          procedure EGR-NGGC-0153 to verify that the licensee incorporates the
Environmental Qualification com) lied with the requirements of 10 CFR
          above comments and clarifications. The reference to a " frozen" drawing
50.49 and 10 CFR 50. Appendix
          to obtain accident temperature data and the wording inconsistencies
3.
          discussed above were identificd to the licensee as a weakness.
An IFI was identified to review
    El.2 Review of Instrument Setooiit Calculations
procedure EGR-NGGC-0153 to verify that the licensee incorporates the
      a.   Insoection Stone (37550)                             ,
above comments and clarifications.
          The inspectors reviewed randomly selected instrument setpoint
The reference to a " frozen" drawing
          calculations to deternine the adequacy of the licensee's calculations.
to obtain accident temperature data and the wording inconsistencies
      b.   Observations and Finninos
discussed above were identificd to the licensee as a weakness.
          The inspectors reviewed the instrument setpoint calculations
El.2 Review of Instrument Setooiit Calculations
          listed below and verified that the calculations were completed in
a.
          accordance with NRC requirements. The inspectors verified that
Insoection Stone (37550),
          the calculations incorporated industry standards. Updated Final
The inspectors reviewed randomly selected instrument setpoint
          Safety Analysis Report commitments. Technical S)ecification
calculations to deternine the adequacy of the licensee's calculations.
          requirements, and recommendations contained in iRC Regulatory
b.
          Guides.   Calculations reviewed were as follows:
Observations and Finninos
          -
The inspectors reviewed the instrument setpoint calculations
                  -Calculation OE41-0036. Power Uprate HPCI Steamline Flow High
listed below and verified that the calculations were completed in
                  Uncertainty and Scaling Calculation.
accordance with NRC requirements. The inspectors verified that
            -
the calculations incorporated industry standards. Updated Final
                  Calculation ORWCU-0010. U1/U2 RWCU Flow Accuracy
Safety Analysis Report commitments. Technical S)ecification
                  Calculation. Units 1 and 2 RWCU Differential Flow Leak
requirements, and recommendations contained in iRC Regulatory
                  Detection / BESS I&C.
Guides.
            -
Calculations reviewed were as follows:
                  Calculation 0821-0068. Power Uprate Main Steam Line Flow
-
                  High Setpoint Uncertainty and Scaling Calculation.
-Calculation OE41-0036. Power Uprate HPCI Steamline Flow High
            -
Uncertainty and Scaling Calculation.
                  Calculation 0-01534A-297. Insulation Resistance Degradation
Calculation ORWCU-0010. U1/U2 RWCU Flow Accuracy
                  Calculation.
-
          From review of System Description SD-01.2. Reactor Vessel
Calculation. Units 1 and 2 RWCU Differential Flow Leak
            Instrumentation. and the Safety Evaluation by the Office of
Detection / BESS I&C.
                                                          ,                                                   _ _ _ _ _ _ _ _ _   a
-
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
20
  Nuclear Reactor Regulation. Conformance to Regulatory Guide 1.97
Nuclear Reactor Regulation. Conformance to Regulatory Guide 1.97
  Revision 2. Brunswick Steam Electric Plant. Units 1 and 2. Dated
Revision 2. Brunswick Steam Electric Plant. Units 1 and 2. Dated
  May 14. 1985. the inspectors concluded that these calculations
May 14. 1985. the inspectors concluded that these calculations
  were. typical. The instrument setpoint calculations typically
were. typical. The instrument setpoint calculations typically
  considered 140 F as the maximum temperature in the calculations.
considered 140 F as the maximum temperature in the calculations.
  From review of the calculations, the inspectors determined that
From review of the calculations, the inspectors determined that
  instruments that perform a safety function are analyzed for a LOCA
instruments that perform a safety function are analyzed for a LOCA
  environment in the reactor building. The calculations showed that
environment in the reactor building.
  instrument uncertainties considered instrument temperature effects
The calculations showed that
  for a maximum temperature of 140' F which is bounding for the
instrument uncertainties considered instrument temperature effects
  analyzed LOCA environment.
for a maximum temperature of 140' F which is bounding for the
  The inspectors also determined that instruments relied upon to
analyzed LOCA environment.
  mitigate the effects of a high energy line break (HELB) were also
The inspectors also determined that instruments relied upon to
  evaluated by the licensee. For this instrumentation,
mitigate the effects of a high energy line break (HELB) were also
  environmental uncertainties-for a harsh environment were not
evaluated by the licensee. For this instrumentation,
  required to be considered since the instrumentation function would
environmental uncertainties-for a harsh environment were not
  occur before the reactor building temperature )rofiles listed in
required to be considered since the instrumentation function would
p the Reactor Building Environmental Report (REBR) Revision 6.
occur before the reactor building temperature )rofiles listed in
  dated November 5. 1997, would reach 140 F and affect instrument
p
  performance. The ins)ectors noted that abnormal temperatures were
the Reactor Building Environmental Report (REBR) Revision 6.
  not discussed in the-RBER. Discussions with licensee engineers
dated November 5. 1997, would reach 140 F and affect instrument
  disclosed that the design base accident event is based on an
performance. The ins)ectors noted that abnormal temperatures were
  initial building environment airspace temperature of 104 F. The
not discussed in the-RBER.
  building temperatures ace measured and recorded daily by plant
Discussions with licensee engineers
  operators in accordance with procedure numbers 101-03.4.1 and 201-
disclosed that the design base accident event is based on an
  03.4,4. Unit 1 and 2 Control Operator Daily Check Sheets.. The
initial building environment airspace temperature of 104
  = operators are required to contact the duty engineer when the
F.
  reactor building temperature exceeds 104 F so that engineering
The
  can perform an assessment of the effects of temperature on
building temperatures ace measured and recorded daily by plant
  environmental qualification.
operators in accordance with procedure numbers 101-03.4.1 and 201-
  -The inspectors noted that calculations for instrumentation which
03.4,4. Unit 1 and 2 Control Operator Daily Check Sheets.. The
  mitigates a HELB demonstrated that the instrument and associated
= operators are required to contact the duty engineer when the
  equipment would not be exposed to a harsh environment before the
reactor building temperature exceeds 104 F so that engineering
    instrumentation performed its safety function. In the instrument
can perform an assessment of the effects of temperature on
  calculations reviewed by the inspectors instrument setpoints were
environmental qualification.
  based on a maximum temperature of 140 F (non-steam environment).
-The inspectors noted that calculations for instrumentation which
  Although allowances were not made for a harsh environment. a
mitigates a HELB demonstrated that the instrument and associated
    seismic allowance was included in the calculations.
equipment would not be exposed to a harsh environment before the
    Review of the temperature profiles as shown in the Brunswick
instrumentation performed its safety function.
    Reactor Building Environmental Report showed that the actuation
In the instrument
    isolation signal would occur before exceeding the temperature
calculations reviewed by the inspectors instrument setpoints were
    allowances assumed in the setpoint uncertainty calculations. An
based on a maximum temperature of 140 F (non-steam environment).
    exce) tion was the High Pressure Coolant Injection (HPCI) line
Although allowances were not made for a harsh environment. a
    breat in the steam tunnel where the temperature profile showed
seismic allowance was included in the calculations.
    that 140 F would be exceeded for ap3roximately 2.5 seconds before
Review of the temperature profiles as shown in the Brunswick
    the isolation trip _ signal occurs. iowever this instrumentation
Reactor Building Environmental Report showed that the actuation
    would remain operable based on thermal delays. However, the HPCI
isolation signal would occur before exceeding the temperature
    isolation function would most likely be initiated by 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
21
                  sensors in the steam tunnel or HPCI room which would occur
sensors in the steam tunnel or HPCI room which would occur
                  imediately with no time delay.
imediately with no time delay.
                  The inspectors concluded that the instrument setpoint calculations
The inspectors concluded that the instrument setpoint calculations
                  complied with NRC requirements and were technically adequate.
complied with NRC requirements and were technically adequate.
                  Review of the calculations showed that environmental effects,
Review of the calculations showed that environmental effects,
j-               specifically accident temperature, were correctly evaluated in the
j-
                  calculations,
specifically accident temperature, were correctly evaluated in the
            c.   Conclusions
calculations,
                  The inspectors concluded that the licensee's calculations were
c.
                  technically adequate and complied with NRC requirements.   The
Conclusions
                  inspectors concurred with the licensee's conclusions that the
The inspectors concluded that the licensee's calculations were
                  setpoints for instruments relied upon to mitigate the effects of a
technically adequate and complied with NRC requirements.
                  KLB did not require inclusion of uncertainties for a harsh
The
                  environment since the instruments perform their ft..iction before
inspectors concurred with the licensee's conclusions that the
                  being effected by the harsh environment. Setpoints for
setpoints for instruments relied upon to mitigate the effects of a
                  instruments required for LOCA effects include the appropriate
KLB did not require inclusion of uncertainties for a harsh
                  environmental uncertainties.
environment since the instruments perform their ft..iction before
        -El.3 Enaineerina Service Reaucst (ESR) 97-00426
being effected by the harsh environment. Setpoints for
            a.   Inspection Scoce (375501                                                   '
instruments required for LOCA effects include the appropriate
                  The inspectors reviewed ESR 97-00426 which was prepared to address
environmental uncertainties.
                  questions on instrument setpoints.
-El.3 Enaineerina Service Reaucst (ESR) 97-00426
            b.   Observations and Findinas
a.
                  A review of procedures and various documents by an independent
Inspection Scoce (375501
                  consultant resulted in questions involving environmental effects
'
                  including uncertainties on instrument accuracy. These guestions         .
The inspectors reviewed ESR 97-00426 which was prepared to address
                  were dccumented in an E-mail message dated June 20, 1997 Subject:         '
questions on instrument setpoints.
                  E0 and Instrument Accuracy. The licensee addressed the referenced         !
b.
                  memo in Engineering Service Request ESR 97-00426. Revision 0.
Observations and Findinas
                -dated September 18. 1997. ESR 97-00426 documents the evaluation
A review of procedures and various documents by an independent
                  completed by the licensee to address environmental effects on-
consultant resulted in questions involving environmental effects
                  instrumentation. The inspectors noted that the licensee response
including uncertainties on instrument accuracy. These guestions
                  did not address the questions in the June 20, 1997 E-mail message
were dccumented in an E-mail message dated June 20, 1997 Subject:
                  point by point. but provided an evaluation that was more generic
.'
                  in nature. The inspectors noted that ESR 97-00426 was an
E0 and Instrument Accuracy. The licensee addressed the referenced
                  engineering disposition (ED) type ESR. as defined in procedure
!
                  EGR-NGGC-0005.   The use of this type ESR to respond to the E-mail
memo in Engineering Service Request ESR 97-00426. Revision 0.
                  cuestions was appropriate since the ESR only communicated existing
-dated September 18. 1997.
                  cesign requirements, did not produce design output, and did not
ESR 97-00426 documents the evaluation
                  change existing engineering documents.
completed by the licensee to address environmental effects on-
                  The ESR concluded that instruments that aerform a safety function
instrumentation.
                  are analyzed for a LOCA environment in t1e reactor building. The
The inspectors noted that the licensee response
                  instrument uncertainties consider -instrument temperature ef fects
did not address the questions in the June 20, 1997 E-mail message
                    for a maximum temperature of 140"F which is the maximum bounding
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
                                                                                          o
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
22
            revised. as discussed below.
temperature for the analyzed LOCA environment.
            ESR 97-00426 also concluded that harsh environmental effects have
The inspectors
            been appropriately accounted for in safety related uncertainty
noted that the word minimum had been incorrectly used in the
            calculations. The ESR concluded that the isolaticr. aquence for a
fourth line, third paragraph in Section 2.0 of the ESR. The
            HELB due to main steam line break. reactor core isolation cooling
licensee stated that they will correct this error when the ESR is
l           steam-line break, high pressure coolant injection steam line
revised. as discussed below.
            break, cr a piping failure in the reactor water cleanup system is
ESR 97-00426 also concluded that harsh environmental effects have
            such thtt the isolation function will occur before the
been appropriately accounted for in safety related uncertainty
            instrumentation is exposed to harsh environmental effects. This
calculations. The ESR concluded that the isolaticr. aquence for a
            conclusion was based on the instrumentation being able to perform
HELB due to main steam line break. reactor core isolation cooling
            its safety function prior to the temperature exceeding the
l
            temperature allowance assumed in the setpoint calculations. For
steam-line break, high pressure coolant injection steam line
            area temperatures exceeding the setpoint temperature uncertainty
break, cr a piping failure in the reactor water cleanup system is
            allowance, the use of emergency operating procedures (EOPs),
such thtt the isolation function will occur before the
            operator action, and local temperature instrumentation would
instrumentation is exposed to harsh environmental effects.
            mitigate the event and provide the actions to determine and/or
This
            maintain. reactor level during a LOCA or HELB.
conclusion was based on the instrumentation being able to perform
            When temperatures exceed the temperatures (140 F) assumed in the
its safety function prior to the temperature exceeding the
            setpoint calculations, plant operation is controlled through the
temperature allowance assumed in the setpoint calculations.
          ' COPS.   A review'of E0P-03-SCCP Revision 5. Secondary Containment
For
            Control Procedure, and 2EOP-LPC Revision 1. Level / Power Control,
area temperatures exceeding the setpoint temperature uncertainty
            shows that high area temperatures are an entry condition into
allowance, the use of emergency operating procedures (EOPs),
            secondary containment control procedure E0P when area temperatures
operator action, and local temperature instrumentation would
            exceed the maximum safe operating value requiring manual reactor
mitigate the event and provide the actions to determine and/or
            sCrdm.
maintain. reactor level during a LOCA or HELB.
            E0P-03-SCCP Revision 5. refers the operators to Caution 1 to
When temperatures exceed the temperatures (140 F) assumed in the
            determine reactor level instrumentation operability. A review of
setpoint calculations, plant operation is controlled through the
            Caution 1 disclosed that vessel level wide range instrumentation                         ;
' COPS.
            8B21 - LI - R604A/604B and C32 - PR - R609 are not to be used when
A review'of E0P-03-SCCP Revision 5. Secondary Containment
              secondary containment temperature exceeds 140 F. This exclusion
Control Procedure, and 2EOP-LPC Revision 1. Level / Power Control,
            was because the reference leg and associated instrumentation for
shows that high area temperatures are an entry condition into
              these loops are in secondary containment. E0P Caution 1 then
secondary containment control procedure E0P when area temperatures
              )rovided compensation data for the remaining level instrumentation
exceed the maximum safe operating value requiring manual reactor
              ]ased on drywell tem]erature, reactor saturation limit, and
sCrdm.
              reactor pressure.         iowever, for secondary containment
E0P-03-SCCP Revision 5.
              temperatures above 140 F. Caution 1 instrumentation may not be
refers the operators to Caution 1 to
              o)erable with instrumentation exposed to temperatures greater
determine reactor level instrumentation operability. A review of
              tlan 140*F during an event. In cases when vessel level can not
Caution 1 disclosed that vessel level wide range instrumentation
              adequately be determined, the E0Ps direct the operators to
;
              depressurize by initiating ADS and flood the vessel using low
8B21 - LI - R604A/604B and C32 - PR - R609 are not to be used when
              pressure emergency core cooling systems.
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
23
                                    the questions in the June 20. 1997 E-mail message regarding
c.
                                    instrument and E0 accuracy. However, the licensee stated that
Conclusions
    _
The inspectors concluded that the licensee adequately addressed
                                    they will revise F.SR 97 00426 to address each question and
the questions in the June 20. 1997 E-mail message regarding
                                    recommendation ir. the E-mail message point by point to further
instrument and E0 accuracy.
                                    clarify their response to the concerns / issues raised in the
However, the licensee stated that
                                    June 20, 1997 E mail message.
_
            El.4 Environmental Qualificat%1
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)
,
,
              a.                      Insnection Scooe (37550.92903)
The inspectors reviewed the licensee's corrective actions for the
                                    The inspectors reviewed the licensee's corrective actions for the
:
:
                                    Environmental Qualification (FO) program, in response to findings
Environmental Qualification (FO) program, in response to findings
l                                     identified during Self-Assessment numbers 95-0041 and 96-0271 and
l
                                    the violations identified in NRC IR 50-325(324)/96-14.
identified during Self-Assessment numbers 95-0041 and 96-0271 and
              b.                   Observations and Findinas
the violations identified in NRC IR 50-325(324)/96-14.
                                      1) Review of E0 Equipment Data Base
b.
                                    The licensee's corrective actions to resolve the discrepancies in
Observations and Findinas
                                    the E0 program identified by NRC (See IR 50-325 324/96-14)
1) Review of E0 Equipment Data Base
                                      include corrections to and updating of the Equipment Data Base
The licensee's corrective actions to resolve the discrepancies in
                                    System (EDBS). Numerous errors in EDBS had been identified and
the E0 program identified by NRC (See IR 50-325 324/96-14)
                                    corrected by the licensee since the inspection findings were
include corrections to and updating of the Equipment Data Base
                                      identified in IR 50-325(324)/96-14. The errors in EDBS were                     .
System (EDBS). Numerous errors in EDBS had been identified and
                                      identified during E0 equipment walkdowns and review of various                   !
corrected by the licensee since the inspection findings were
                                      data bases. In addition, numerous errors were identified in the
identified in IR 50-325(324)/96-14. The errors in EDBS were
                                      EQ zones listed in EDBS for the location where various components
.
                                    were installed. These primarily occurred at. zone boundaries and
identified during E0 equipment walkdowns and review of various
                                      were being resolved during review of walkdown data.
!
                                      The requirements for. recording and correcting E0 data in EDBS was
data bases.
                                      s)ecified in- CP&L procedures EGR-NGGC-0156 and ENP-33,6.   The
In addition, numerous errors were identified in the
                                  -c1anges to EDBS to correct errors were processed using Form 100 of
EQ zones listed in EDBS for the location where various components
                                      ENP-33.6. The Form 100 was design verified in the E0 unit and was
were installed.
                                      then forwarded to appropriate personnel for entry into EDBS. All
These primarily occurred at. zone boundaries and
                                      EDBS data entries made were independently verified by personnel in
were being resolved during review of walkdown data.
                                      the Configuration Management group in the Design Control Unit.
The requirements for. recording and correcting E0 data in EDBS was
                                      The independent verification was performed to minimize o-
s)ecified in- CP&L procedures EGR-NGGC-0156 and ENP-33,6.
                                      eliminate data entry errors. Additional corrections to EDBS were
The
                                      ongoing to incorporate E0 walkdown ins)ection results and the
-c1anges to EDBS to correct errors were processed using Form 100 of
                                      revisions to EQ qualification data paccages.
ENP-33.6. The Form 100 was design verified in the E0 unit and was
                                      The inspectors reviewed some randomly selected revisions to EDBS
then forwarded to appropriate personnel for entry into EDBS. All
                                      identified as a result of the E0 corrective actions and verified
EDBS data entries made were independently verified by personnel in
                                      the EDBS data had been corrected. The inspectors also discussed
the Configuration Management group in the Design Control Unit.
                                      the program for control of changes to EDBS with various licensee
The independent verification was performed to minimize o-
                                      personnel who perform the day to day system revisions. These
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
24
  discussions disclosed that these individuals were cognizant of the
discussions disclosed that these individuals were cognizant of the
  requirements for controlling and making corrections to EDBS.
requirements for controlling and making corrections to EDBS.
  2) Review of Qualification Data Packages
2) Review of Qualification Data Packages
  The inspectors reviewed a draft cop.f of Revision 4 of ODP No. 49.
The inspectors reviewed a draft cop.f of Revision 4 of ODP No. 49.
  titled. " Qualification Data Package For NAMCO EA180 Series Limit
titled. " Qualification Data Package For NAMCO EA180 Series Limit
  Switches" to determine if it adequately demonstrated environmental
Switches" to determine if it adequately demonstrated environmental
  qualification for the safety related NAMCO switches for use inside
qualification for the safety related NAMCO switches for use inside
  the drywell in accordance with 10 CFR 50.49 and appropriate
the drywell in accordance with 10 CFR 50.49 and appropriate
  licensee E0 Prccedures. The package addressed the following:
licensee E0 Prccedures.
  qualification level (0588 Cat. I); tag numbers of equipment
The package addressed the following:
  covered in the QDP: test report aaplicability; similarity of                           test
qualification level (0588 Cat. I); tag numbers of equipment
  specimens to installed equipment: E0 parameters. temperature,
covered in the QDP:
  pressure, relative humidity, radiation, chemical spray,
test report aaplicability; similarity of test
  submergence; cualified life: E0 maintenance requirements; test
specimens to installed equipment:
  anomalies; anc operating experience items.
E0 parameters. temperature,
  During review 3f the Draft ODP. the inspectors identified the following
pressure, relative humidity, radiation, chemical spray,
  questions / comments:
submergence; cualified life:
  .
E0 maintenance requirements; test
          The text in the CDP indicates that there were five anomalies in
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
'
'
          Qualification Test Report (OTR) 130 but only four anomalies were
discussed in the ODP.
          discussed in the ODP.
l
l .
Attachment 2 to the ODP included a calculation for qualified life
          Attachment 2 to the ODP included a calculation for qualified life
.
l         of the limit switches which was not signed as reviewed.
l
  *        Differences were noted in the system component evaluation
of the limit switches which was not signed as reviewed.
          worksheets (SCEW) for the same limit switches in the different
Differences were noted in the system component evaluation
          units.
*
  *        Data was missing from some of the SCEW sheets. That is, there
worksheets (SCEW) for the same limit switches in the different
          were blanks on the data sheets.                       For example, data on accuracy was
units.
          left blank.
Data was missing from some of the SCEW sheets.
  *
That is, there
          Some components were specified with Anaconda flex and others just
*
          stainless steel flex conduit. Additionally, only certain
were blanks on the data sheets.
          components were specified for weep holes.
For example, data on accuracy was
  *      Page 49 section 4.1 Installation requirements indicates that the
left blank.
          conduit seal may not be necessary for those limit switches
Some components were specified with Anaconda flex and others just
          installed in the Reactor Building. This requirement should be
*
          clear and should specifically list those limit switches which
stainless steel flex conduit.
          require conduit selling to ensure qualification.
Additionally, only certain
  *      Page 13 lists the 16 Namco EA180 limit switches which had been
components were specified for weep holes.
          installed. However only 14 were considered qualifieo by this ODP.
Page 49 section 4.1 Installation requirements indicates that the
          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
conduit seal may not be necessary for those limit switches
          equivalent switches were not discussed in the ODP.
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
25
        maintenance practice to lubricate the NAMCO limit switches.
In Section 2 of the 00P it was stated that it was a good
        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
maintenance practice to lubricate the NAMCO limit switches.
        refurbished. However, a statement was made on page 21 that
however. lubrication was not specified in Section 4 of the ODP
        qualified replacement part kits were no longer available.
which lists recommended maintenance practices.
*      A reference was made to abnormal temperatures on page 38 of the
In Section 4.2 of the ODP it was stated that the switches can be
        ODP.   However, abnormal temperatures were not included in DR 227.
*
*      The inspectors questioned apparent inconsistencies between
refurbished. However, a statement was made on page 21 that
        activation energies and aging methods discussed in referenced
qualified replacement part kits were no longer available.
        qualification test reports (OTRs).
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 licensee indicated that these comments would be evaluated by
the E0 group and if appropriate, addressed in Revision 4 of the
the E0 group and if appropriate, addressed in Revision 4 of the
Line 1,454: Line 1,904:
The inspectors reviewed a draft copy of Revision 7 of ODP-67
The inspectors reviewed a draft copy of Revision 7 of ODP-67
General Electric Company IC 7700 Series Motor Control Centers for
General Electric Company IC 7700 Series Motor Control Centers for
BNP.   The GE MCCs. located or, the 20, 50, and 80 foot elevations
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
of the Units 1 and 2 Reactor Buildings, are subject to harsh
environments resulting from postulated design basis accidents and         ;
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.
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
A series of similarity analysis were performed to demonstrate
similarity between the tested configuration and supplied. The
similarity between the tested configuration and supplied.
inspectors reviewed portions of DR 232. "Nutherm Report No. CPL-
The
inspectors reviewed portions of DR 232. "Nutherm Report No. CPL-
7806R. Qualification Test Results Applicable to Brunswick Nuclear
7806R. Qualification Test Results Applicable to Brunswick Nuclear
Power Plant Safety-Related GE 7700 MCCs." Revision 0, dated
Power Plant Safety-Related GE 7700 MCCs." Revision 0, dated
June 30, 1997 which dccumented the similarity analysis. Section 2 of
June 30, 1997 which dccumented the similarity analysis.
Section 2 of
DR-232 contains a discussion on the similarity analysis between the
DR-232 contains a discussion on the similarity analysis between the
components tested by NUTHERM and those installed in the Brunswick MCCs.
components tested by NUTHERM and those installed in the Brunswick MCCs.
The similarity discussion covers fuses, stab assemblies, control
The similarity discussion covers fuses, stab assemblies, control
transformers control and power wiring, overload heaters. overload
transformers control and power wiring, overload heaters. overload
relays, terminal boards, starters and contactors, molded case circuit
relays, terminal boards, starters and contactors, molded case circuit
breakers. circuit protectors. disconnect switches. potentiometers, and
breakers. circuit protectors. disconnect switches. potentiometers, and
indicating lights. The similarity analyses were based on the similarity
indicating lights. The similarity analyses were based on the similarity
analyses contained in DR 1.1. GE Company NEDC-30696-P. May 1985. MCC
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
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
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
require a similarity analysis.
that NOTHERM was able to establish that the com3onents they tested were
Based on review of DR-232 NRC concluded
  in the same family as those provided by GE in t1e MCCs. This review was
that NOTHERM was able to establish that the com3onents they tested were
also dccumented in IR 50-325(324)/97-09.
in the same family as those provided by GE in t1e MCCs. This review was
A draft copy of Revision 0 of ODP 99. R. G. Laurence Series 500
also dccumented in IR 50-325(324)/97-09.
and 600 Solenoid Valves was reviewed. The inspectors verified
A draft copy of Revision 0 of ODP 99. R. G. Laurence Series 500
that similarity analysis was included in the ODPs.
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
26
    reactor building. Tha E0 walkdowns were performed in accordance
3) Review of EO Walkdown Data
    with CP&L Special Procedure OSP-96-014. EQ Equipment Field
The inspectors reviewed E0 walkdown data which document inspection
    Verification. The pyrpose of the walkdowns was to verify the
of E0 equipment-in the Unit 2 MSIV pit and drywell, and Unit 2
    accuracy of the manulacturer/model number listed in the licensee's
reactor building.
    data bases and to verify the equipment installed orientation and
Tha E0 walkdowns were performed in accordance
    configuration were in accordance with the E0 qualification
with CP&L Special Procedure OSP-96-014. EQ Equipment Field
    documentation.         The ins)ectors reviewed walkdown records for scram                     '
Verification. The pyrpose of the walkdowns was to verify the
    pilot' solenoid valves, 1AMC0 limit switches, temperature elements,
accuracy of the manulacturer/model number listed in the licensee's
    excess flow check valves, and pressure switches. The walkdown
data bases and to verify the equipment installed orientation and
    data was recorded on field inspection data sheets which were'then
configuration were in accordance with the E0 qualification
    converted into an electronic data base. The inspectors verified
documentation.
    that discrepancies identified during the walkdowns were documented
The ins)ectors reviewed walkdown records for scram
    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-
pilot' solenoid valves, 1AMC0 limit switches, temperature elements,
    AF JR1, 97-AFUR2, 97- A UR3, and 97-AFUR4. These WR/J0s document
excess flow check valves, and pressure switches. The walkdown
    drilling of weepholes in junction boxes in the Unit 2 MSIV pit to
data was recorded on field inspection data sheets which were'then
    resolve a moisture intrusion issue. These boxes are associated
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.
with limit switches for the Unit 2 main steam isolation valves.
L   The completed WR/J0s showed that the weepholes were drilled to
L
    resolve the concerns. The inspectors did not identify any
The completed WR/J0s showed that the weepholes were drilled to
;  discrepancies in the records reviewed.
resolve the concerns. The inspectors did not identify any
    4) Review of Environmental Qualification Condition Reports
discrepancies in the records reviewed.
    The inspectors reviewed the licensee's corrective c.,ctions to
;
L   disposition the CRs listed below. These CRs were initiated by the
4) Review of Environmental Qualification Condition Reports
    licensee to-document and disposition nonconforming items whicn
The inspectors reviewed the licensee's corrective c.,ctions to
    were identified during the ongoing E0 reconstitution project. The
L
    nonconforming items were identified as a result of E0 equipment
disposition the CRs listed below.
    walk h ns, review cnd updating of E0 equipment ODPs, omissions
These CRs were initiated by the
    from the original program, or changes to the operating
licensee to-document and disposition nonconforming items whicn
    environment. The CRs reviewed were as follows:
were identified during the ongoing E0 reconstitution project.
    CR 97-02015
The
    The licensee initiated CR 97-02015 on June 6. 1997 to document and
nonconforming items were identified as a result of E0 equipment
    disposition deficiencies that had been identified by the
walk h ns, review cnd updating of E0 equipment ODPs, omissions
    licensee's training staff during observation of simulator training
from the original program, or changes to the operating
    when the fire protection system had not been isolated within the
environment. The CRs reviewed were as follows:
    15 minute time period after initiation of a HELB specified in
CR 97-02015
    31 ant o)erating 3rocedures. The 15 minute time period is the
The licensee initiated CR 97-02015 on June 6. 1997 to document and
    ) asis w1ich esta)lished flood '.evels for E0 e
disposition deficiencies that had been identified by the
    and north and south RHR and core spray rooms.quipment               in the HPCI
licensee's training staff during observation of simulator training
                                                            Review of closure
when the fire protection system had not been isolated within the
    for CR 97-02015 disclosed that the licensee concluded that the
15 minute time period after initiation of a HELB specified in
    issue has been adequately addressed by operator training,
31 ant o)erating 3rocedures. The 15 minute time period is the
    primarily through critiques which were held following the
) asis w1ich esta)lished flood '.evels for E0 e
    completion of the simulator training to discuss deficiencies noted
and north and south RHR and core spray rooms.quipment in the HPCI
    during the training. In response to the CR. Action Items were
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
Y
                                                                                  _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ _


  - -. - - -           -         __ - . - - - - .
- -. - - -
              .
-
__ - . - - - - .
.
1
1
                                                      27
27
assigned to the Operator Training group to incorporate the basis
'
'
                assigned to the Operator Training group to incorporate the basis
for the need to isolate the fire protection system into training
,
,
                for the need to isolate the fire protection system into training
materials.
                materials. However, review of the training records on June 12,
However, review of the training records on June 12,
                1997, by personnel from the E0 group resulted in additional
1997, by personnel from the E0 group resulted in additional
                questions regarding the licensee s corrective actions. The
questions regarding the licensee s corrective actions. The
                records reviewed by the E0 personnel indicated that during
records reviewed by the E0 personnel indicated that during
simulator training, approximately 10 to 20 percent of the
'
'
                simulator training, approximately 10 to 20 percent of the
operators were failing to enter AOP-05,0, Radioactive Spills. High
                operators were failing to enter AOP-05,0, Radioactive Spills. High
Radiation, and Airborne Activity, or were entering the AOP late
.                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
(after 15 minutes). The inspectors made an indepen6nt review of
                disclosed that the records the E0 personnel reviewed on June 12,
the training records reviewed by the E0 personnel.
                1997 were for the six month
This review
                02015 (January - June 1997) The    .
,
                                                      period prior to reviewed
disclosed that the records the E0 personnel reviewed on June 12,
                                                          inspectors  initiation of CR 97-
1997 were for the six month
                                                                                training
02015 (January - June 1997) period prior to initiation of CR 97-
                records for July - September, 1997 and noted significant
The inspectors reviewed training
                improvement in this area, although the HELB scenario was not
.
                included as part of the simulator training exercises in this time
records for July - September, 1997 and noted significant
                period. The training scenario did include a torus leak which
improvement in this area, although the HELB scenario was not
                required entry into A0P-05.0.
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
I
                The inspectors noted that the concern regarding flooding of
The inspectors noted that the concern regarding flooding of
_
_
                  instruments could also be caused by other accidents such as pipe
instruments could also be caused by other accidents such as pipe
L               breaks in the service water or Reactor Building Closed Cooling
L
l               Water (RBCCW). Operator actions in these cases would be directed
breaks in the service water or Reactor Building Closed Cooling
e                by E0P-03 SCCP Secondary Containment Control Proccdure (SCCP),
l
                based on high water leve'is in the HPCI and north and south RHR and
Water (RBCCW). Operator actions in these cases would be directed
                core spray rooms. An uttry into E0P-03-SCCP would also result
by E0P-03 SCCP Secondary Containment Control Proccdure (SCCP),
                  from flooding in these same rooms caused by activation of the fire
e
                protection system. As aaditional followup on this issue, the
based on high water leve'is in the HPCI and north and south RHR and
                inspectors observed simulator training scenarios performed on
core spray rooms. An uttry into E0P-03-SCCP would also result
                December 3 and 17, 1997. Included in the scenario was a RCIC
from flooding in these same rooms caused by activation of the fire
                steam line break (HELB) and activitation of the fire protection
protection system.
                system. Both crews participating in the training scenario
As aaditional followup on this issue, the
                isolated the fire protection system within the 15 minute time
inspectors observed simulator training scenarios performed on
                period. The inspectors also questioned some randomly selected
December 3 and 17, 1997.
                reactor operators regarding the need for entry into A0P-05.0
Included in the scenario was a RCIC
                following a HELB. The operators were cognizant of the basis of
steam line break (HELB) and activitation of the fire protection
                the actions in A0P-05.0 (need and reason for isolating the ' ire
system.
                protection s
Both crews participating in the training scenario
                CR 97-02015.ystem) and were familiar with the problem addrc ses by
isolated the fire protection system within the 15 minute time
                The inspectors verified the action items associated with the CR
period.
                were completed.       CR 97-02015 was closed on December 11. 1997.
The inspectors also questioned some randomly selected
                CR 97-01841. 97 02025. & 97-02408 These CRs documented various
reactor operators regarding the need for entry into A0P-05.0
                issues regarding possible effects of moisture on E0 equipment. CR
following a HELB. The operators were cognizant of the basis of
                97-0184) was initiated to document the effect of spray from the
the actions in A0P-05.0 (need and reason for isolating the ' ire
                fire protection system on E0 equipment in the reactor building.
CR 97-02015.ystem) and were familiar with the problem addrc ses by
                The licensee has resolved all the issues associated with this CR
protection s
                except for drilling of weepholes in junction boxes whicn may be
The inspectors verified the action items associated with the CR
                affected by the water spray. Licensee engineers are currently
were completed.
                - preparing instructions and procedures for completing this work.
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
28
    The problem documented in CR 97-02025 concerned an issue which had
The problem documented in CR 97-02025 concerned an issue which had
    been the subject of IE Circular 79-05. Moisture Leakage in
been the subject of IE Circular 79-05. Moisture Leakage in
    Stranded Wire Conductors, which was issued by NRC on March 20.
Stranded Wire Conductors, which was issued by NRC on March 20.
    1979. This affects Patel seals which were used to seal some
1979.
    stranded wire conductors in instrument circuits. CR 97-0?408
This affects Patel seals which were used to seal some
    documents several other moisture intrusion issues. The immediate
stranded wire conductors in instrument circuits.
    corrective action taken to resolve these issues, as documented in
CR 97-0?408
    CR 97-02408 was to hire an outside consultant to address the
documents several other moisture intrusion issues.
    issues. The consultant has reviewed many of the issues documented
The immediate
    in CR numbers 97-01841, 97-02025. and 97-02408 and made
corrective action taken to resolve these issues, as documented in
    recommendations, some of which have been implemen.ed. The
CR 97-02408 was to hire an outside consultant to address the
    consultant also addressed another issue in the CRs involving
issues.
    current leakage in control circuit and the possible impact on ODPs
The consultant has reviewed many of the issues documented
    and E0 of equipment. This concern was the effect of moisture
in CR numbers 97-01841, 97-02025. and 97-02408 and made
    intrusion through stranded wire conductors, sealed with Patel
recommendations, some of which have been implemen.ed.
    seals, which could result in leakage currents in instrument
The
    circuits. ESR 97 00440 was issued for the 120 volt AC circuits and
consultant also addressed another issue in the CRs involving
    ESR 97-00441 for DC circuits. These ESRs are currently being
current leakage in control circuit and the possible impact on ODPs
    reviewed by licensee engineers. The current leakage issue was
and E0 of equipment. This concern was the effect of moisture
    also applicable to questions raised regarding the NAMCO limit
intrusion through stranded wire conductors, sealed with Patel
    switches. The inspectors will review the licensee's evaluation of
seals, which could result in leakage currents in instrument
    current leakage and its ap311 cation to evaluation of E0 equipment
circuits. ESR 97 00440 was issued for the 120 volt AC circuits and
    in a future inspection. T11s was identified to the licensee as
ESR 97-00441 for DC circuits.
    IFl 50 325(324)/97-13-07. Review Technical Evaluation of Current
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
!
!
    Leakage and the Effect on EQ Equipment. pending further review by
NRC.
    NRC.
The licensee also aerformed an evaluation of the potential for
    The licensee also aerformed an evaluation of the potential for
moisture wicking t1 rough Patel seals.
    moisture wicking t1 rough Patel seals. This evaluation was
This evaluation was
i   documented in ESR 97-00423. 03erability Evaluation - Wicking.
i
    Review of the ESR disclosed t.at the licensee performed a detailed
documented in ESR 97-00423. 03erability Evaluation - Wicking.
    evaluation of the Patel seals by comparison of the installations
Review of the ESR disclosed t.at the licensee performed a detailed
    at Brunswick with the configurations tested by NRC at Sandia
evaluation of the Patel seals by comparison of the installations
    Laboratorics (NUREG/CR 0699. Jublished March.1979).                 The
at Brunswick with the configurations tested by NRC at Sandia
    licensee's conclusions were t1at the design function of the
Laboratorics (NUREG/CR 0699.
    instellea equipment will not be effected by moisture intrusion
Jublished March.1979).
    through the stranded wire. The ESR was based on a review of the
The
    duration of the design accidents and the resulting leakage
licensee's conclusions were t1at the design function of the
    currents caused by moisture intrusion into limit switches.
instellea equipment will not be effected by moisture intrusion
    Further review of this ESR will be performed as part of IFI 50-325
through the stranded wire.
    (324)/97-13-07, discussed above.
The ESR was based on a review of the
    CR 97-02016 & 97-02074
duration of the design accidents and the resulting leakage
    CR numbers 97 02016 & 97-02074 were initiated to document issues
currents caused by moisture intrusion into limit switches.
    involving NAMCO limit switches. The following issues were
Further review of this ESR will be performed as part of IFI 50-325
    identified in the CRs:
(324)/97-13-07, discussed above.
    *      Inability to identify the date of manufacture of the switches
CR 97-02016 & 97-02074
            since the codes for date of manufacture were painted over.
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
I
                                              29
29
  *      Potential for paint to impair the operability of the switches.
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.
The concern was that paint on the roller arms would impair
mechanical function of the switches.
'
'
  *      Chemical reaction between paint and internal switch components
Chemical reaction between paint and internal switch components
        would cause corrosion of switches, leading to failure of the
*
        switches.
would cause corrosion of switches, leading to failure of the
  *      Use of incorrect qualification test reports (0TRs) in the
switches.
        qualification test reports which qualified the switches.
Use of incorrect qualification test reports (0TRs) in the
  *      Effect of current leakage on switch operability.
*
  A total of 14 NAMCO limit switches were covered under the E0
qualification test reports which qualified the switches.
  program. These switches were installed during modifications
Effect of current leakage on switch operability.
  completed in 1983 and 1984 The licensee has determined that none
*
  of the switches were purchased or manufactured prior to 1980.
A total of 14 NAMCO limit switches were covered under the E0
  Therefore, the concern raised by IE Bulletin 79-28. Possible
program.
  Malfunction of NAMC0 Model EA 180 Limit Switches at Elevated
These switches were installed during modifications
  Temperatures, would not apply to the switches installed at
completed in 1983 and 1984
  Brunswick, Review of the licensee's response to IEB 79-28
The licensee has determined that none
  disclosed that none of the potentially defective switches had been
of the switches were purchased or manufactured prior to 1980.
  purchased by the Brunswick site.
Therefore, the concern raised by IE Bulletin 79-28. Possible
  Review of the i1censee's corrective actions completed to date
Malfunction of NAMC0 Model EA 180 Limit Switches at Elevated
  disclosed that the following actions have been completed:
Temperatures, would not apply to the switches installed at
  *
Brunswick,
        The licensee has identified the date of manufacture for most of
Review of the licensee's response to IEB 79-28
        the NAMCO limit switches.             Additional manufacture dates may be
disclosed that none of the potentially defective switches had been
        identified when the Unit 1 walkdowns are completed during the
purchased by the Brunswick site.
        Spring 1998 refueling outage. However, the licensee has
Review of the i1censee's corrective actions completed to date
        conclusively determined that none of the switches would be
disclosed that the following actions have been completed:
        affected by the defects identified in IEB 79-28.
The licensee has identified the date of manufacture for most of
  .
*
        The switches were stroked in accordance with frequencies per the
the NAMCO limit switches.
        Technical Specifications which demonstrates that the mechanical
Additional manufacture dates may be
        function of the switches had not been impaired by the paint.
identified when the Unit 1 walkdowns are completed during the
  *      The paint has been tested.               The test results show the
Spring 1998 refueling outage.
        not cause corrosion or deterioration of the switches paint would
However, the licensee has
  .      The ODP. has been revised to incorporate the correct OTRs. The
conclusively determined that none of the switches would be
        ODP. ODP 49, was still in draft.
affected by the defects identified in IEB 79-28.
  .      The current leakage issue has been evaluated " ESR numbers 97-
The switches were stroked in accordance with frequencies per the
        00440 and 97-00441, which are currently being reviewed by licensee
.
        engineers.
Technical Specifications which demonstrates that the mechanical
  The licensee subsequently has determined that the switches were
function of the switches had not been impaired by the paint.
  still within their qualified ;'fe. No equiament operability
The paint has been tested.
  issues related to tv.e NAMCO ilmit switches lave been identified.
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
30
  [R 97 02367
[R 97 02367
  This CR was initiated on July 3, 1997 to document the failure to
This CR was initiated on July 3, 1997 to document the failure to
  initiate CRs for nonconforming items, specifically, MCC door
initiate CRs for nonconforming items, specifically, MCC door
  gaskets and non standard Raychem splices identified as a
gaskets and non standard Raychem splices identified as a
  violation by NRC during an inspection documented in NRC 1R 50-325
violation by NRC during an inspection documented in NRC 1R 50-325
  (324)/97 08.- The licensee's corrective actions included
(324)/97 08.- The licensee's corrective actions included
  completion of a review of all the E0 walkdown data sheets to
completion of a review of all the E0 walkdown data sheets to
  identify any nonconforming equipment. Additional corrective
identify any nonconforming equipment.
  actions included training of personnel in the E0 group regarding
Additional corrective
  the corrective action program and assessment of the effectiveness
actions included training of personnel in the E0 group regarding
  of the corrective actions. These correcthe actions were also
the corrective action program and assessment of the effectiveness
  associated with other similar corrective action CRs. such as CR
of the corrective actions.
  97 01972 and CR 97-02465. The inspectors reviewed the completed
These correcthe actions were also
  corrective actions and concurred with closure of CR 97 02367. The
associated with other similar corrective action CRs. such as CR
  CR was closed on December 14. 1997.
97 01972 and CR 97-02465.
  CR 97-02465 and 97-02672
The inspectors reviewed the completed
  This CR wac initiated on July 15, 1997, to document concerns on EQ
corrective actions and concurred with closure of CR 97 02367.
  operability determinations.     This CR referenced CR numbers 97-
The
  01841, 97 02025. and 97 02408. discussed above, which involve
CR was closed on December 14. 1997.
  moisture intrusion issues. As a result of the concerns raised in
CR 97-02465 and 97-02672
  CR 97 02465, the E0 group presented an action plan to resolve the
This CR wac initiated on July 15, 1997, to document concerns on EQ
  moisture intrusion issues (CR 97 02465) to the plant nuclear
operability determinations.
  safety committee. Although, further review showed the operability
This CR referenced CR numbers 97-
  determinations for the three CRs were correct, the root cause
01841, 97 02025. and 97 02408. discussed above, which involve
  analysis concluded that there were other problems which resulted
moisture intrusion issues.
  in CR 97-02465.
As a result of the concerns raised in
  The root cause of CR 97-02465 was attributed-to weak E0 project
CR 97 02465, the E0 group presented an action plan to resolve the
  management. The root cause/ event review for the CR listed the
moisture intrusion issues (CR 97 02465) to the plant nuclear
  causal-factors indicative of weak E0 3roject management to be poor
safety committee. Although, further review showed the operability
  communications within the E0 group, tie site position that E0
determinations for the three CRs were correct, the root cause
  problems were primarily docunitation problems, and a poor
analysis concluded that there were other problems which resulted
  corrective action culture within the E0 group. The poor
in CR 97-02465.
  corrective action culture was evidenced by corrective action items
The root cause of CR 97-02465 was attributed-to weak E0 project
  which were routinely extended, overdue, or completed late: failure
management.
  to prepare JCOs: numerous CRs written against the E0 grou) for
The root cause/ event review for the CR listed the
  improper corrective actions: and closing CR action items )y other
causal-factors indicative of weak E0 3roject management to be poor
  action items without completing the corrective actions. A
communications within the E0 group, tie site position that E0
  violation of NRC requirements was identified in IR 50 325, 324/97-
problems were primarily docunitation problems, and a poor
  12 for failure of the licensee to implement their corrective
corrective action culture within the E0 group.
  action program.
The poor
  The licensee's corrective actions to address the issues raised in
corrective action culture was evidenced by corrective action items
  CR 97-02465 included increased management oversight     aerforming a
which were routinely extended, overdue, or completed late: failure
  review of the E0 project schedule to complete the higlest priority
to prepare JCOs: numerous CRs written against the E0 grou) for
  work activities first, conducting more frequent E0 group meetings
improper corrective actions: and closing CR action items )y other
  to improve communications within the E0 group, transferring some
action items without completing the corrective actions. A
  E0 group functions from the Design Control l%1t to a site
violation of NRC requirements was identified in IR 50 325, 324/97-
  organization. and performance of an effer' ve. dss review of the
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
31
              completed corrective actions. The CR was closed on December 17,
completed corrective actions.
              1997. The inspectors reviewed the completed corrective actions
The CR was closed on December 17,
              .and concurred with closure of the CR. The ins)ectors concurred
1997.
              with the licensee's conclusions that the opera]ility
The inspectors reviewed the completed corrective actions
              determinations for the three referenced CRs were appropriate. NRC
.and concurred with closure of the CR.
              will perform review of the liccasee's actions to correct the         l
The ins)ectors concurred
              violation in future inspections,
with the licensee's conclusions that the opera]ility
              CR 97-02672, which was inniated on August 5. 1997, indicated that
determinations for the three referenced CRs were appropriate. NRC
              the Supervisor comments listed in CR 97 02465 were a misstatement
will perform review of the liccasee's actions to correct the
              of the consensus of opinion of individuals which met to discuss CR
l
              97-02465. Review of CR 97-02672 disclosed that the CR did not
violation in future inspections,
              raise any new issues or conceriis which had not been addressed by
CR 97-02672, which was inniated on August 5. 1997, indicated that
              CR 97 02465. CR 97-02672 was closed on December 17, 1997.     NRC
the Supervisor comments listed in CR 97 02465 were a misstatement
              concurs with the licensee's conclusions and closure of the CR.
of the consensus of opinion of individuals which met to discuss CR
              CR 97 4059
97-02465. Review of CR 97-02672 disclosed that the CR did not
              This CR was initiated on December 2, 1997, to document concerns
raise any new issues or conceriis which had not been addressed by
              and questions on ESR 97-00426. The questions involved
CR 97 02465. CR 97-02672 was closed on December 17, 1997.
              appropriateness of E0P actions, the need to include evaluation of
NRC
              drywell instrumentation in tic ESR, and various questions on
concurs with the licensee's conclusions and closure of the CR.
              instrument setpoints. The 1 Lensee completed a review of the
CR 97 4059
              questions raised in the CR and concluded that the ESR had
This CR was initiated on December 2, 1997, to document concerns
              addressed these issues, or the issues were beyond the scope of the
and questions on ESR 97-00426.
              ESR,   For exam)le, appropriateness of E0P actions were approved by
The questions involved
              NRC for all BW1s and do not involve instrument setpoints. There
appropriateness of E0P actions, the need to include evaluation of
              are no instruments in the drywell which provide signals for
drywell instrumentation in tic ESR, and various questions on
              automatic actuation. The inspectors reviewed the licensee's
instrument setpoints.
              responses to the questions in the CR and concurred with the
The 1 Lensee completed a review of the
              licensee's conclusions that no new corrective actions were
questions raised in the CR and concluded that the ESR had
              required to resolve the concerns / questions raised in CR 97-04059-
addressed these issues, or the issues were beyond the scope of the
              which had not been previously resolved.
ESR,
              5) Review of Environmental Qualification Requirements in
For exam)le, appropriateness of E0P actions were approved by
              Procurement Practices
NRC for all BW1s and do not involve instrument setpoints.
              Th'e inspectors reviewed CP&L procedure MCP-NGGC-0401, Material
There
              Acquisition (Procurement. Receiving, and Shipping). Revision 4,
are no instruments in the drywell which provide signals for
              dated August 26, 1997. This procedure specifies the instructions
automatic actuation.
              for procurement of safety related materials for use in CP&L
The inspectors reviewed the licensee's
              nuclear plant. The inspectors noted that the requirements for
responses to the questions in the CR and concurred with the
              obtaining reviews by E0 engineers is specified in the procedure.
licensee's conclusions that no new corrective actions were
              Discussions with licensee engineers and review of previous
required to resolve the concerns / questions raised in CR 97-04059-
              revisions of the procedure disclosed that the procedure had been
which had not been previously resolved.
              revised to strengthen the need for the E0 review in Revision 2 of
5) Review of Environmental Qualification Requirements in
              .MCP-NGGC 0401, effective April 15. 1997. Revision 2 added
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
requirements that components that require environmental
              qualification:be reviewed by the E0 group.
qualification:be reviewed by the E0 group.
              During review of CRs. the inspectors identified several examples
During review of CRs. the inspectors identified several examples
              of acceptance of materials / equipment by procurement engineering
of acceptance of materials / equipment by procurement engineering


                  _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
                                                            32
32
  for use in E0 installations which were based on test reports which ;
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
had not been reviewed by the E0 group -These were documented in
  discrepancies in documents prepared by procurement engineering
'CR numbers 97-01970 and 97-03036,
  which affected E0 equipment were also identified during review of
Several additional examples of
  procurement specifications and other documents su * as material
discrepancies in documents prepared by procurement engineering
  evaluations. These discrepancies were documented in CR 97-04035
which affected E0 equipment were also identified during review of
  which tas initiated on November 25, 1997. The review of
procurement specifications and other documents su * as material
  procurement documents was being performed as part of the
evaluations.
  corrective actions to address the E0 program discrepancies
These discrepancies were documented in CR 97-04035
  identified in IR 50-325(324)/96 14. This was listed as Commitment   *
which tas initiated on November 25, 1997. The review of
  #4 in the licensee's December 19, 1996 Reply to Notice of           ,
procurement documents was being performed as part of the
  Violation,
corrective actions to address the E0 program discrepancies
  6) Equipment Lubrication Requirements
identified in IR 50-325(324)/96 14.
  The inspectors reviewed CP&L procedure MMM-053. Equipment
This was listed as Commitment
  Lubrication Application Guidance and Lubricant Listing,
*
  Revision 6 dated November 11, 1997. This procedure provides a
#4 in the licensee's December 19, 1996 Reply to Notice of
  listing of plant equipment with recommended lubricants to be used,
,
  guidelines for lubrication of plant equipment, and lubricant
Violation,
  sampling methods. The inspectors identified the following issues
6) Equipment Lubrication Requirements
  after reviewing the procedure:
The inspectors reviewed CP&L procedure MMM-053. Equipment
  -
Lubrication Application Guidance and Lubricant Listing,
          ODPs 26, 68, and 88 were not referenced in procedure MMM-
Revision 6 dated November 11, 1997.
          053. These ODPs cover environmental qualification of
This procedure provides a
          Reliance electric motors.
listing of plant equipment with recommended lubricants to be used,
  -      Document References corresponding to above ODPs were not
guidelines for lubrication of plant equipment, and lubricant
          referenced.
sampling methods.
  -      The types of lubricant specified fo, the Reliance motors in
The inspectors identified the following issues
          procedure MMM 053 differ from those listed in the ODPs 26
after reviewing the procedure:
          and 68.
ODPs 26, 68, and 88 were not referenced in procedure MMM-
  -
-
          Procedure MMM 053 permits maintenance to change the
053.
          lubricant without obtaining engineering review or approval.
These ODPs cover environmental qualification of
  Discussions with licensee engineers disclosed the CR 97-04015 was
Reliance electric motors.
  initiated on November 20, 1997, to document the fact that the
Document References corresponding to above ODPs were not
  procedure permits changes to lubricants without performance. of an
-
  engineering review. Action Item 40 to CR 97-02627 was issued to
referenced.
  document a similar issue. This action item was closed by CR 97-
The types of lubricant specified fo, the Reliance motors in
  04015.
-
  The inspectors determined that the licensee had not evaluated that
procedure MMM 053 differ from those listed in the ODPs 26
  the type of lubricants (Mobil) specified-in procedure MMM 053 for
and 68.
  Reliance electric motors differed from those listed in ODP 26 and
Procedure MMM 053 permits maintenance to change the
  68, Review of ODP 26. Revision 1. Joy Fan /Peliance Electric
-
  Company, Class 1E Continuous Duty, 20 HP, and ODP 68, Revision 5.
lubricant without obtaining engineering review or approval.
  -Standby Gas Treatment System - Fair Company Filter Unit and
Discussions with licensee engineers disclosed the CR 97-04015 was
  Control, showed that the electric motors were both qualification
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
33
'
'
          tested using Chevron SRI 2 grease.         The impact of using a
tested using Chevron SRI 2 grease.
          dif ferent type of grease to lubricate the motors on the
The impact of using a
          environmental qualification testing of the motors had not been
dif ferent type of grease to lubricate the motors on the
          documented by the licensee. The licensee initiated CR 97 04064 to
environmental qualification testing of the motors had not been
          document the fact that substitution of alternate lubricants had
documented by the licensee.
          not been evaluated by E0 engineers. The failure to establish
The licensee initiated CR 97 04064 to
          maintenance procedures appropriate to the circumstances for
document the fact that substitution of alternate lubricants had
          performing maintenance was identified to the licensee as another
not been evaluated by E0 engineers.
          example of violation item 50 325(324)/97-13-02. Inadequate
The failure to establish
          Procedure for the Conduct of E0 Preventive Maintenance.
maintenance procedures appropriate to the circumstances for
      c. Conclusions
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
1
          One violation example was identified regarding an inadequate E0
One violation example was identified regarding an inadequate E0
          maintenance procedure for lubrication of E0 electric motors. Two
maintenance procedure for lubrication of E0 electric motors.
          inspector followup items were identified to followu) on revisions
Two
          to instrument setpoint procedures and to review leacage current
inspector followup items were identified to followu) on revisions
          calculations. The licensee was making progress in resolving and
to instrument setpoint procedures and to review leacage current
          closing CRt identified by the E0 group. As of the inspection
calculations.
          dates, no 0DPs had been issued.
The licensee was making progress in resolving and
    E5   Engineering staff Knowledge and Qualification
closing CRt identified by the E0 group.
    E5.1 Trainino and Qualification of E0 Personnel
As of the inspection
      a. Insnection Scone (37550)
dates, no 0DPs had been issued.
          The inspector reviewed the licensee's program for training and
E5
          qualification of personnel in the E0 task force. including both
Engineering staff Knowledge and Qualification
          CP&L and contract engineers,
E5.1 Trainino and Qualification of E0 Personnel
      b. Observations and Findinos
a.
          The requirements for performance of E0 equipment walkdowns are
Insnection Scone (37550)
          specified in CP&L Special Procedure OSP-96-014. E0 Equipment Field
The inspector reviewed the licensee's program for training and
          Verification. The prerequisite in procedure OSP 96-014 for
qualification of personnel in the E0 task force. including both
          individuals performing the walkdowns was to read the procedure.
CP&L and contract engineers,
          The licensee qualified a number of individuals to perform the
b.
          field walkdowns through a training program conducted in accordance
Observations and Findinos
          with CP&L procedure TI-100. Conduct of Training. These
The requirements for performance of E0 equipment walkdowns are
          individuals included Instrumentation and Control technicians.
specified in CP&L Special Procedure OSP-96-014. E0 Equipment Field
          contract engineers, and personnel assigned to the E0 group who
Verification.
          were qualified E0 engineers. The training for the qualified E0
The prerequisite in procedure OSP 96-014 for
          engineers consisted of reading the procedure. orientation and on-
individuals performing the walkdowns was to read the procedure.
          the-job training to become familiar with the walkdown and data
The licensee qualified a number of individuals to perform the
s        gathering process. For other personnel, the training included
field walkdowns through a training program conducted in accordance
          reading of the procedures, formal classroom lectures.
with CP&L procedure TI-100. Conduct of Training.
          demonstrations, performance of practical exercises, and on-the-job
These
          training. The walkdown group supervisor performed a detailed
individuals included Instrumentation and Control technicians.
          review of the result < of practical exercises and data gathered   <
contract engineers, and personnel assigned to the E0 group who
          during initial walke is prior to signifying the individuals were
were qualified E0 engineers.
                                                              s. _       __
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
gathering process.
For other personnel, the training included
s
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
34
  qualified to perform walkdowns. The training provided for the
qualified to perform walkdowns.
  walkdcwn personnel exceeded the procedural requirements. The E0
The training provided for the
  walkdown grou) supervisor stated that the level of training
walkdcwn personnel exceeded the procedural requirements.
  provided to t1e walkdown personnel war to assure that the walkdown
The E0
  results were very accurate and to preclude the need for repeat
walkdown grou) supervisor stated that the level of training
  work. The inspectors revieweJ the training records for the
provided to t1e walkdown personnel war to assure that the walkdown
  walkdown personnel and verified that they had been trained in
results were very accurate and to preclude the need for repeat
  accordance with the licensee's program. The inspectors noted that
work.
  the experience level for the walkdown personnel varied from a
The inspectors revieweJ the training records for the
  recent graduate engineer to individuals with more than 20 years of
walkdown personnel and verified that they had been trained in
  experience. The inspectors reviewed the walkdown inspection
accordance with the licensee's program.
  records prepared by various individuals in the walkdown group and
The inspectors noted that
  noted that the original walkdown records were complete and
the experience level for the walkdown personnel varied from a
  accurate, with some exceptions. Discussions with the walkdown
recent graduate engineer to individuals with more than 20 years of
  group supervisor disclosed that corrections noted on the records
experience.
  were the result of reviews perfnrmed to resolve discrepancies in
The inspectors reviewed the walkdown inspection
  the records. The changes were made as a result of additional
records prepared by various individuals in the walkdown group and
  walkdown inspections which were doc'mented in the records. In one
noted that the original walkdown records were complete and
  case, an individual was terminated for failure to perform the
accurate, with some exceptions.
  walkdowns and complete the walkdown records properly.   This
Discussions with the walkdown
  individual's work was reviewed by the licensee and corrected where
group supervisor disclosed that corrections noted on the records
  necessary.
were the result of reviews perfnrmed to resolve discrepancies in
  The inspectors also reviewed the training and qualification
the records.
  records for E0 technical personnel. These records included
The changes were made as a result of additional
  previous work experience, education and training, and CP&L
walkdown inspections which were doc'mented in the records.
  specific training applicable to the E0 project. This training
In one
  included E0 technical reviewer, E0 design verifier E0
case, an individual was terminated for failure to perform the
  calculations, and E0 ESR originator.   The inspectors also
walkdowns and complete the walkdown records properly.
  questioned the manager of the E0 group concerning work assignments
This
  within the E0 grou). That is, assignment of specific activities
individual's work was reviewed by the licensee and corrected where
  to individuals wit 1 previous experience in a particular area of
necessary.
  specialization, such as review of requirements for qualification
The inspectors also reviewed the training and qualification
  of motors or specific types of instrumentation. The E0 group
records for E0 technical personnel.
  manager has recently are)ared a directory of all engineers working
These records included
  within the E0 group w11c1 lists each engineer's experience and
previous work experience, education and training, and CP&L
  what work activities they have completed for the E0 project at
specific training applicable to the E0 project.
  Brunswick. The purpose of this directory was for the engineers
This training
  within the group to know who has worked on various problems and
included E0 technical reviewer, E0 design verifier E0
  issues so they could obtain assistance from these individuals when
calculations, and E0 ESR originator.
  they become involved with similar technical issues. The directory
The inspectors also
  was distributed'to personnel in the E0 group. The E0 group
questioned the manager of the E0 group concerning work assignments
  manager provided a copy of the directory to the inspectors and
within the E0 grou). That is, assignment of specific activities
  discussed the basis for the various work assignments within the
to individuals wit 1 previous experience in a particular area of
  group.which were based on the past work experience of the E0
specialization, such as review of requirements for qualification
  technical personnel,
of motors or specific types of instrumentation. The E0 group
c. Conclusions
manager has recently are)ared a directory of all engineers working
  The inspector concluded that the licensee's program for training
within the E0 group w11c1 lists each engineer's experience and
  and qualification of E0 engineers meets NRC requirements.
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
35
        E8   Hiscellaneous Engineering Issues (37551, 92903)
E8
        E8.1   (Closed) URI 50-325(324)/97-08-04: Control of Ecuioment Data Base
Hiscellaneous Engineering Issues (37551, 92903)
              System (EDBS) Information
E8.1
              The licensee issued CR 97-02400. Non Validated EDBS Information,
(Closed) URI 50-325(324)/97-08-04: Control of Ecuioment Data Base
              concerning rc, tine use of non-validated EDBS information. This wes
System (EDBS) Information
              associated with VIO 50 325(324)/97-08 03. Safety Relay Setting Change
The licensee issued CR 97-02400. Non Validated EDBS Information,
              Made as Pen and Ink Changes to Procedure. The licensee replied to this
concerning rc, tine use of non-validated EDBS information.
              violation on September 2. 1997. The reply discussed licensee corrective
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
'
'
              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
              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
              Safe Shutdown Impairment. addressed corrective action fcr use of an EDBS
non validated field for determination of an Alternate Safe Shutdown
              non validated field for determination of an Alternate Safe Shutdown
impairment.
              impairment.     Plant procedure OENP-33.6. Equipment Data Base System
Plant procedure OENP-33.6. Equipment Data Base System
              Control and Revision, provides instructions for control of EDBS
Control and Revision, provides instructions for control of EDBS
              information. Color coding of fields in the electronic database
information. Color coding of fields in the electronic database
              represent the various types of data present. This procedure provides
represent the various types of data present.
              direction that certain types of data are not to be used until verified.
This procedure provides
              Accordingly two previous violations address the use of non-verified
direction that certain types of data are not to be used until verified.
              EDBS information. The licensee corrective actions for these violations
Accordingly two previous violations address the use of non-verified
              are being completed. The requirements for the control of information
EDBS information.
              are in procedure OENP-33.6. Previous items address the concern of this
The licensee corrective actions for these violations
              URI. therefore this item is closed.
are being completed.
        E8.2 (Closed) LER 50-325(324)/97-04:       Soent Fuel Shionina Cask Handlina
The requirements for the control of information
              Activities
are in procedure OENP-33.6.
              This report documented the discovery by the licensee that the heavy load
Previous items address the concern of this
              analysis as described in tne UFSAR did not completely bound movement of
URI. therefore this item is closed.
              the shiroing cask from the primary lift to the secondary lift with the
E8.2 (Closed) LER 50-325(324)/97-04:
              valve box covers removed.       It was determined that movement of the cask
Soent Fuel Shionina Cask Handlina
              with a non single failure proof yoke and less than full cask integr'ty
Activities
              constituted an unreviewed safety question (US0) in accordance with the
This report documented the discovery by the licensee that the heavy load
              requirements specified in 10 CFR Part 50.55         The failure to obtain
analysis as described in tne UFSAR did not completely bound movement of
              prior approval for a previously unanalyzed condition was determined in
the shiroing cask from the primary lift to the secondary lift with the
              IR 50-325(324)/97-12 to be a violation. In a letter to the NRC dated
valve box covers removed.
              August 6. 1997, the licensee requested a license amendment for review of
It was determined that movement of the cask
              the US"     The licensee re evaluated findings relative to the 30 foot
with a non single failure proof yoke and less than full cask integr'ty
              dro: ~cident and qualified the transfer yoke using guidance provided in
constituted an unreviewed safety question (US0) in accordance with the
              NUR b 0612. Control of Heavy Loads at Nuclear Power Plants. This
requirements specified in 10 CFR Part 50.55
              evaluation contended that a fuel shipping cask drop event was not
The failure to obtain
              credible. therefore operation with less than full cask integrity was no
prior approval for a previously unanalyzed condition was determined in
              longer a problem due to acceptable redundancy in the lifting yoke. In a
IR 50-325(324)/97-12 to be a violation.
              letter to the licensee dated December 2. 1997, the NRC accepted the
In a letter to the NRC dated
              licensee determination that operation with the valve covers removed
August 6. 1997, the licensee requested a license amendment for review of
              would not compromise the health and safety of the public due to
the US"
              acceotable redundancy of the lift devices. Based on the acceptance by
The licensee re evaluated findings relative to the 30 foot
              the NRC of the licensee's evaluation and issuance of the enforcement
dro: ~cident and qualified the transfer yoke using guidance provided in
              action as described in IR 50-325(324)/97-12 this item is closed.
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
4
                                                                                            36
36
                          E8.3 (Closed) Inspector Followun item 50-325(324)/96-14-05.'Effect of EO
E8.3 (Closed) Inspector Followun item 50-325(324)/96-14-05.'Effect of EO
                                          Accuracy on Instrument Setooint Calculations.
Accuracy on Instrument Setooint Calculations.
'
'
                                            Review of procedures and various documents by an independent                                             *
Review of procedures and various documents by an independent
                                            contJ1 tant had resulted in a number of questions regarding the
contJ1 tant had resulted in a number of questions regarding the
*
~
effect of environmental effects (uncertainties) on instrument
;
;
~
accuracy
                                            effect of environmental effects (uncertainties) on instrument
The questions / concerns were documented in an E mail
                                            accuracy        The questions / concerns were documented in an E mail                                   *
*
                                          message dated June 20, 1997. subjert E0 and Instrument Accuracy,
message dated June 20, 1997. subjert E0 and Instrument Accuracy,
                                            in order to address the issues raised in the June 20 E mail
in order to address the issues raised in the June 20 E mail
                                            message,       a review of instrument setpoint calculations was
message,
                                            performed by licensee instrumentation and controls (l&C)
a review of instrument setpoint calculations was
                                            engineers. . The review was documented in ESR 97-000426, which 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
'
'
                                            discussed in paragraph El.3. above. The inspectors also reviewed
various instrument setpoint calculations (documented in paragraph
                                            various instrument setpoint calculations (documented in paragraph
E1.2. above) and determined that E0 accuracy has been aroperly
"
"
                                            E1.2. above) and determined that E0 accuracy has been aroperly
considered in the instrument setpoint calculations.
                                            considered in the instrument setpoint calculations. T1e
T1e
                                            ins)ectors had no further questions regarding instrument setpoint
ins)ectors had no further questions regarding instrument setpoint
                                            metloaology or accuracy at this time.
metloaology or accuracy at this time.
                          E8.4             JClosed) Violation item 50-325(324)/97-02-08. Failure to Imolement ao
E8.4
                                              nsoection Procram for Safety-Related Miscellaneous Structural Steel
JClosed) Violation item 50-325(324)/97-02-08. Failure to Imolement ao
                                            The licensee responded to this violation in letters dated
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
April 30. 1997, and June 26. 1997 Subject: Reply to Notice of
                                            Specification 248-107 and review of other specifications to assure OC
Violation.
                                            inspection criteria required by applicable codes and standards
The licensee's corrective actions included revision of
                                            referenced in the UFSAR had been included in the specifications.
Specification 248-107 and review of other specifications to assure OC
                                            Specifications reviewed included the following: 248-117 - Installation
inspection criteria required by applicable codes and standards
                                            of Piping Systems: 048 012 - Installation of Electrical Cables: 006 001
referenced in the UFSAR had been included in the specifications.
                                            - Design. Testing & Inspection of Concrete Mixes. Concrete Materials and
Specifications reviewed included the following:
                                          High-Strength Bolts: 005-005 - Design. Testing, & Inspection of Concrete
248-117 - Installation
                                          Mixes. Concrete Materials: 013 001 - Concrete Work: and 018-002.
of Piping Systems: 048 012 - Installation of Electrical Cables: 006 001
                                          Miscellaneous Steel.             Additional corrective actions included inspection
- Design. Testing & Inspection of Concrete Mixes. Concrete Materials and
                                            of a sample of safety related high strength bolts installed using
High-Strength Bolts: 005-005 - Design. Testing, & Inspection of Concrete
                                            Specification 248-107. The inspectors reviewed the results of the
Mixes. Concrete Materials: 013 001 - Concrete Work: and 018-002.
                                            structural steel inspectior.s which were documented in ESR 97-00085.
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
-
-
                                            hiscellaneous Structural Steel Connection Inspections. The licensee
also revised procedure MMP-013. to incorporate the specification 248-107
                                            also revised procedure MMP-013. to incorporate the specification 248-107
changes and trained OC. engineering and planning personnel on the
                                            changes and trained OC. engineering and planning personnel on the
changes to specification 248-107 which now require additional QC
                                            changes to specification 248-107 which now require additional QC
inspections.
                                            inspections. The inspectors reviewed records which documented
The inspectors reviewed records which documented
                                            inspections performed for selected USl A-46 modifications completcd on
inspections performed for selected USl A-46 modifications completcd on
                                            Unit 1 during the Fall.1997 refueling outage and verified the
Unit 1 during the Fall.1997 refueling outage and verified the
                                            structural steel inspections were completed in accordance with the
structural steel inspections were completed in accordance with the
                                            revised procedures.
revised procedures.
4
4
                          Ee.5 (Closed) Violation item 50-325(324)/97 08-07. Failure to initiate
Ee.5 (Closed) Violation item 50-325(324)/97 08-07. Failure to initiate
                                            Condition Reports to [,0cument Nonconformina E0 Items
Condition Reports to [,0cument Nonconformina E0 Items
I
I
  .                                        The licensee
The licensee reshonded to this violation in a letter dated
                                            September     2. 199 reshonded
September 2. 199
                                                                          Subject: Reply to this  violation
Subject: Reply to Notice of Violation.
                                                                                              to Notice of Violation. in a letter    The dated
The
.
,
,
          _ _ ,                         _ . - . . _ . _ ,     . . . _ _           - , . -       _                                 -       -_ __,
_ _ ,
_ . - . . _ . _ ,
.
. . . _ _
.
c
- , . -
_
-
-_
__,


                                        37
37
      licensee's corrective actions included training of E0 personnel on
licensee's corrective actions included training of E0 personnel on
      the corrective action program, a review of che E0 walkdown data
the corrective action program, a review of che E0 walkdown data
      sheets to identify any potential nonconforming conditions which
sheets to identify any potential nonconforming conditions which
      had not been previously identified and dispositioned, and
had not been previously identified and dispositioned, and
      organizational changes to improve management o"ersight in the E0
organizational changes to improve management o"ersight in the E0
      group. CR 97-02367 was initiated by the licensee on July 3. 1997
group.
      to document and disposition the two s)ecific examples of failure
CR 97-02367 was initiated by the licensee on July 3. 1997
      to initiate CRs identified by NRC. Tie inspectors ceviewed the CR
to document and disposition the two s)ecific examples of failure
      closecut records (CR was closed on December 14, 1997) and the
to initiate CRs identified by NRC.
      licensee's corrective actions and verified that the actions were
Tie inspectors ceviewed the CR
      completed in accordance with the licensee's violation response.
closecut records (CR was closed on December 14, 1997) and the
                                IV. Plant SuppEt
licensee's corrective actions and verified that the actions were
R1   Radiological Protection and Chemistry Controls
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
RI.1 Use of locks to Control Access
  a. Insnection Stone (71750)
a.
      The inspector verified a selected sampling of doors required to be
Insnection Stone (71750)
      locked, by plant TSs and procedures, fc r the purpose of radiation
The inspector verified a selected sampling of doors required to be
      protection,
locked, by plant TSs and procedures, fc r the purpose of radiation
  b. Observations and Findinas
protection,
      The inspector reviewed Environmental & Radiological Control 0E&RC-0040.
b.
      Control of Locked High Radiation and Very High Radiation Areas, to
Observations and Findinas
      determine the controls used to lock high radiation area doors and
The inspector reviewed Environmental & Radiological Control 0E&RC-0040.
      barriers. The inspector located a sampling of the locked high radiation
Control of Locked High Radiation and Very High Radiation Areas, to
      area doors specified in OE&RC-0040 and tested them to ensure that they
determine the controls used to lock high radiation area doors and
      were locked. The ins)ector found that all the locked high radiation
barriers.
      doors tested were locced,
The inspector located a sampling of the locked high radiation
  c. Conclusions
area doors specified in OE&RC-0040 and tested them to ensure that they
      The ins)ector determined that each of the locked high radiation area
were locked. The ins)ector found that all the locked high radiation
      dcors w11ch were checked were locked.   The ins)ector concluded that the
doors tested were locced,
      licensee is satisfactorily controll1ng locked ligh radiation areas in
c.
      the plant.
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
R1.2 Radioactive Material Controls
  a. insoection Scqoe (71750)
a.
      The inspector conducted a housekeeping tour of radioactive material
insoection Scqoe (71750)
      storage areas located in outside areas within the protected area,
The inspector conducted a housekeeping tour of radioactive material
  b, Observations and Findinas
storage areas located in outside areas within the protected area,
      The inspector found several poor radiological work practices in the
b,
      radiological material (RAM) storage area located aojacent to the
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
38
                  scaffolding identified as radioactive lying unprotected on a wooden
Radiological Maintenance Service Building in the northwest corner of the
                  pa l l e'. .
p.*otected area.
                  The ~icensee conducted a walkdown of this area and the radiological
A bucket containing scaffolding brackets was half
                  service building, and identified multiple conditions requiring action.
filled with water and was labeled as radioactive material.
                  These items were identified in CR 97-04122. Nonconforming Material
The label
                  Condition,
identified the brackets as contaminated.
      c.         Conclusions
This practice had the
                  The inspector determined that several poor radiological work practices
possibility of allowing the potentially contaminated water to cause a
                  existed in a radioactive material storage area.
spread of contamination in an RAM storage area.
    S2           Status of Security Facilities and Equipment
There was also
    c2.1 Plant Access Control and Physical Barriers
scaffolding identified as radioactive lying unprotected on a wooden
      a.         Inspection Scone (71750)
pa l l e'. .
                  The inspector verified the status and condition of the protected area
The ~icensee conducted a walkdown of this area and the radiological
                  fencing,
service building, and identified multiple conditions requiring action.
      b.         Qbser"ations Jnd Findinas
These items were identified in CR 97-04122. Nonconforming Material
                  The inspector performed a walkdown of the protected area fence. The
Condition,
                  fence was inspected for integrity such as corrosion on the posts, gaps
c.
                  in the fence, and general adequacy. The inepector noted no
Conclusions
                  deficiencies,
The inspector determined that several poor radiological work practices
      c.         Conclusions
existed in a radioactive material storage area.
                  The inspector found the status and condition of the protected area fence
S2
                  to be satisfactory.
Status of Security Facilities and Equipment
    F1           Control of Fire Protection Activities
c2.1 Plant Access Control and Physical Barriers
    F1.1 Operability of Fire Protection Facilities and Eauioment
a.
      a.         Ipsoection Scone (64704)
Inspection Scone (71750)
                  The inspector reviewed the operation's fire protection daily impairment
The inspector verified the status and condition of the protected area
                  reports on the facility's fire protection systems and features, and
fencing,
                  inspected these items to determine the performance trends and the
b.
                  material conditions of this equipment.
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
4
4
.
.
:
:
'
'
                                                                    39                                               ,
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:                 ,
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 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.
Fire: Detection System -
                                  The impairment status report provided the licensee with a good means of
3
                                  identifying out-of-service fire protection equipment and provided status
3
1
Fire Suppression System
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,
-
-
                                  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
During the plant tours the inspector noted that the maintenance and
                                  material condition of the fire protection equipment were satisfactory.
material condition of the fire protection equipment were satisfactory.
                            c.     Conclusions
c.
Conclusions
'
'
                                  Correstive maintenance on degraded fire protection systems was
Correstive maintenance on degraded fire protection systems was
                                  accomplished in a. timely manner.>The maintenance and material condition
accomplished in a. timely manner.>The maintenance and material condition
of the fire protection equipment and features were satisfactory.
-
,
,
,
                                  of the fire protection equipment and features were satisfactory.
F2
                                                                                    -
Status of Fire Protection Facilities and Equipment
                                                                                                                      ,
F2.1 E3ssive Fire Barriers
                        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 ~ include penetration seals. wraps, walls. structural member--
Fire barriers are used to
                                  fire resistanticoatings.. doors, dampers. etc. Fire barriers are used to
-prevent the spread of fire and to protect redundant safe shutdown
                                -prevent the spread of fire and to protect redundant safe shutdown
equipment.
                                  equipment.   Laboratory testing of fire barrier materials is done only on
Laboratory testing of fire barrier materials is done only on
                                  a-limited range of test assemblies. In-)lant-installations can vary
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
from the tested configurations. -Under tie provisions of Generic Letter
                                  are permitted to develop engineering evaluations justifying such
(GL) 86-10. Implementation of Fire Protection Requirements, licensees
                                  deviations.
are permitted to develop engineering evaluations justifying such
                                                                                                                      ;
deviations.
  w     -, ,, - . . .         -. - -                                                 .     - - ,.             . -
;
w
-, ,, - . . .
-. - -
. .
. - - - . - . . .
-
. -
.
- - ,.
- .
~ -
. -


                      . . _ _ _ _ _ - _ - _ _ - _ _ _ _ _ _ _ _ _ .
. . _
_ _ _ _ - _ - _ _ - _ _ _ _ _ _ _ _ _ .
'
'
                                                                    40
40
  2.2 Silicone foam Penetration Seals
2.2
    a. Inspection Stone (64704)
Silicone foam Penetration Seals
      The inspector reviewed the fire barrier ,,ilicone foam penetration seal
a.
      design end testing. The inspector compared as-built fire barrier
Inspection Stone (64704)
      silicone foam penetratioh seals to fire endurance test configurations to
The inspector reviewed the fire barrier ,,ilicone foam penetration seal
      verify that the as-built penetration seals reviewed were qualified by
design end testing.
      appropriate fire endurance tests, representative of, and bounded by, the
The inspector compared as-built fire barrier
      design and construction of the fire endurance test specimens. During
silicone foam penetratioh seals to fire endurance test configurations to
      plant walkdowns the inspector observed the installation configurations
verify that the as-built penetration seals reviewed were qualified by
      of selected fire barrier silicone foam 3enetration seals to unfirm that
appropriate fire endurance tests, representative of, and bounded by, the
      the licensee had established an accepta)le design basis for those fire
design and construction of the fire endurance test specimens.
      barriers used to separate safe shutdown functions.
During
    b. Observations and Findinas
plant walkdowns the inspector observed the installation configurations
      The inspector reviewed the fire barrier seal design and testing for six
of selected fire barrier silicone foam 3enetration seals to unfirm that
;      of ten fire barrier silicone foam seal penetrations, Additional reviews
the licensee had established an accepta)le design basis for those fire
I     are documented in NRC 1Rs 50-325(324)/92-31, 93 08. and 93-38.
barriers used to separate safe shutdown functions.
      The inspector reviewed Brunswick Specification No. 118 003, Revision 7.
b.
      Selection and Installation of Fire Barrier Penetration Seals: Corrective
Observations and Findinas
      Maintenance Procedure OCMP-010, Revision 2, Installation of Fire
The inspector reviewed the fire barrier seal design and testing for six
      Barrier, Pressure Boundary Penetration and Water / Moisture Seals: Fire
of ten fire barrier silicone foam seal penetrations, Additional reviews
      Protection Procedure FFP-015. Revision 23, Fire Barrier Penetration Seal
;
      Work Control: Periodic Test OPT-34.6.7.12. Revision 3. Fire Barrier
I
      Penetration Seals: and the Fire Hazards Analysis (FHA) for the location
are documented in NRC 1Rs 50-325(324)/92-31, 93 08. and 93-38.
      and description of fire areas: and assessed the licensee's supporting
The inspector reviewed Brunswick Specification No. 118 003, Revision 7.
      technical justification and any available engineering evaluations for
Selection and Installation of Fire Barrier Penetration Seals: Corrective
      the sampled silicone foam type oenetration seals,
Maintenance Procedure OCMP-010, Revision 2, Installation of Fire
      The inspector's review focused on verifying that the following design
Barrier, Pressure Boundary Penetration and Water / Moisture Seals: Fire
      and installation paramaters for the as-built configurations were
Protection Procedure FFP-015. Revision 23, Fire Barrier Penetration Seal
      adequately bounded and justified by the licensee's engineering
Work Control: Periodic Test OPT-34.6.7.12. Revision 3. Fire Barrier
      evaluations:
Penetration Seals: and the Fire Hazards Analysis (FHA) for the location
      .      penetration opening sizes
and description of fire areas: and assessed the licensee's supporting
      e      thermal mass of penetrating items
technical justification and any available engineering evaluations for
      e     clearances of penetrating items
the sampled silicone foam type oenetration seals,
      e     unexposed surface temperatures
The inspector's review focused on verifying that the following design
      The insoector found that penetration seal field verification
and installation paramaters for the as-built configurations were
      documentation was maintained by the licensee.                   However, the seal
adequately bounded and justified by the licensee's engineering
      installers * qualification and training records were not readily
evaluations:
      available for review. Although the installation and repair procedures
penetration opening sizes
      for penetration seals provided adequate guidance to ensure materials
.
      were installed per design requirements, the inspector could not verify
thermal mass of penetrating items
      that the established surveillance recuirements included vendor
e
      recommendations for inspection and icentification of silicone foam seal
clearances of penetrating items
      aging and shrinkage.
e
unexposed surface temperatures
e
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
41
                      evahated the adequacy of the deviations from a tested fire barrier
The licensee was unable to locate the penetration seal testing
                      contiguration. This does not satisfy the guidar.ce of GL 8610. The
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
licensee stated that industry documentation is available to support
i
i
                      licensee stated that industry documentation is available to support
silicone foam penetration seal installations at Brunswick but the
                      silicone foam penetration seal installations at Brunswick but the
.tiformation was maintained at other Carolina Power and Light (CP&L)
                      .tiformation was maintained at other Carolina Power and Light (CP&L)
sites.
                      sites.
The penetration seal testing documentation, vendor data and inspection
                      The penetration seal testing documentation, vendor data and inspection
criteria, installer qualification and training records, and evaluations
                      criteria, installer qualification and training records, and evaluations
of deviations from tested fire barrier configurations will be reviewed
                      of deviations from tested fire barrier configurations will be reviewed
during a subsequent NRC inspection.
                      during a subsequent NRC inspection. This is identified as IFl 50 325
This is identified as IFl 50 325
                      (324)/97-13 04. Review of Licensee Records and Engineering Evaluations
(324)/97-13 04. Review of Licensee Records and Engineering Evaluations
                      to Establish the Fire Resistant Capabilities of Fire Rated Silicone foam
to Establish the Fire Resistant Capabilities of Fire Rated Silicone foam
                      Penetration Seals,
Penetration Seals,
                  c. Conclusions
c.
                      The inspector concluded that silicone foam penetration seal field
Conclusions
                      verification documentation was maintained by the licensee. The
The inspector concluded that silicone foam penetration seal field
                      installation and repair procedures for penetration seals provided
verification documentation was maintained by the licensee.
                      adequate guidance to ensure that materials were installed per design
The
                      requirements. However, the designs were not supported by seal testing
installation and repair procedures for penetration seals provided
                      documentation, vendor data and inspection criteria, installer
adequate guidance to ensure that materials were installed per design
                      qualification and training records, and engineering evaluations that
requirements.
                      satisfy the guidance of GL 8610 for deviations from the fire barrier
However, the designs were not supported by seal testing
                      configuration qualified by tests.
documentation, vendor data and inspection criteria, installer
                F2.3 Fire Doors
qualification and training records, and engineering evaluations that
                  a. Insnection Scone (64704)
satisfy the guidance of GL 8610 for deviations from the fire barrier
                      The inspector reviewed UFSAR Section 9.5.1.4.3.4.b. Fire Doors, and
configuration qualified by tests.
                      performed plant walkdowns to verify that the UFSAR wording was
F2.3 Fire Doors
                      consistent with the observed plant installation configurations for
a.
                      selected fire doors installed in fire barriers used to separate safe
Insnection Scone (64704)
                      shutdown functions.
The inspector reviewed UFSAR Section 9.5.1.4.3.4.b. Fire Doors, and
                b.   Observations and Findinas
performed plant walkdowns to verify that the UFSAR wording was
                      The UFSAR St.ction 9.5.1.4.3.4.b. Fire Doors, states that doors and
consistent with the observed plant installation configurations for
                      frames are either listed by a national testing laboratory or are
selected fire doors installed in fire barriers used to separate safe
                      constructed similar to listed doors and frames. All doors and frames
shutdown functions.
                      have been evaluated to assure satisfactory ratings. Results are
b.
                      documented in the FHA. During the review of the FHA the inspector
Observations and Findinas
                      identified that, while evaluations of fire doors and frames existed. the
The UFSAR St.ction 9.5.1.4.3.4.b. Fire Doors, states that doors and
                    -licensee failed to document their results in the FHA. which is section
frames are either listed by a national testing laboratory or are
                      9.5.1.5 of the UFSAR.
constructed similar to listed doors and frames.
                                                                                                                                  1
All doors and frames
                                - - ,       r     ,,-                 ~   , - , , - , , - , - - - - - -             -   , -v ,
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
42
                      After discussions with the licensee. CR 97-04103 was issued to track the
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.
l
                        This UFSAR discrepancy was identified by the inspector and is discussed
failure to provide the results of fire door evaluations in the FHA.
                        in Section F2.4.
This UFSAR discrepancy was identified by the inspector and is discussed
                      A review of the surveillance ins)ection and testing procedures for fire
in Section F2.4.
                        doors was performed to confirm tlat the licensee specified fire door
A review of the surveillance ins)ection and testing procedures for fire
                        clearance acce)tance criteria was in accordance with the guidance of
doors was performed to confirm tlat the licensee specified fire door
                        National Fire )rotection Association (NFPA) 80. Standard for Fire Doors
clearance acce)tance criteria was in accordance with the guidance of
                        and Fire Windows. On December 10. 1997. the inspector observed ongoing
National Fire )rotection Association (NFPA) 80. Standard for Fire Doors
                        door replacement and repair activities for fire doors in the DG
and Fire Windows. On December 10. 1997. the inspector observed ongoing
                        building. No discrepancies were identified,
door replacement and repair activities for fire doors in the DG
                    c.   Conclusions
building.
No discrepancies were identified,
c.
Conclusions
I
I
                        The inspector concluded that fire door surveillance prc:edures and
The inspector concluded that fire door surveillance prc:edures and
                        acceptance criteria for verification o' fire daor clearances were in
acceptance criteria for verification o' fire daor clearances were in
                        accordance with NFPA quidance.                   Howevr a UFSAR discrepancy associated
accordance with NFPA quidance.
                        documentation of fire door and frame eu.uations was identified.
Howevr
                F2.4 UFSAR Review
a UFSAR discrepancy associated
                        A recent discovery of a licensee o)erating the facility in a manner
documentation of fire door and frame eu.uations was identified.
                        contrary to the UFSAR description lighlighted the need for a special
F2.4 UFSAR Review
                        focused review that compares plant practices, procedures, and/or
A recent discovery of a licensee o)erating the facility in a manner
                        parameters to the UFSAR descriptions. While performing the inspections
contrary to the UFSAR description lighlighted the need for a special
                        discussed in this report. the inspector reviewed the applicable portions
focused review that compares plant practices, procedures, and/or
                        of the UFSAR that related to the areas inspected. The inspector
parameters to the UFSAR descriptions.
                        verified that the UFSAR wording was consistent with the observed plant
While performing the inspections
                        practices, procedures, and/or parameters.
discussed in this report. the inspector reviewed the applicable portions
                        The inspector reviewed UFSAR Section 9.5.1.4.3.4.b, Fire Doors, as part
of the UFSAR that related to the areas inspected.
                        of the fire protection program review activiti u , An inconsistency was
The inspector
                        noted in that the licensee failed to document the results of evaluations
verified that the UFSAR wording was consistent with the observed plant
                        of fire doors and frames in the FHA which is section 9.5.1.5 of the
practices, procedures, and/or parameters.
                        UFSAR.                 This issue is discussed in Section F2.3. This item will be
The inspector reviewed UFSAR Section 9.5.1.4.3.4.b, Fire Doors, as part
                        identified as part of URI 50-325(324)/97-13-05. UFSAR Discrepancy Fire
of the fire protection program review activiti u ,
                        Doors.
An inconsistency was
                F3     Fire Protection Procedures and Documentation
noted in that the licensee failed to document the results of evaluations
                F3.1 Fire Protection Procedures
of fire doors and frames in the FHA which is section 9.5.1.5 of the
                    a.   Insoection Scone (64704)
UFSAR.
                        The inspector evaluated the adequacy and implementation of the
This issue is discussed in Section F2.3.
                        licensee s Eire Protection Program described in the UFSAR and in Plant
This item will be
                        Operating Manual Fire Protection Procedure OPLP 01. Revision 6. Fire
identified as part of URI 50-325(324)/97-13-05. UFSAR Discrepancy Fire
                        Protection Program Document. In addition a comparison was made of the
Doors.
                        program to selected NRC Safety Evaluation Reports which ap3 roved the
F3
                        station fire protection program. The inspector reviewed t7e following
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:
>
>
                        procedures for compliance with the NRC requirements and guidelines:
.
    .
.
  .  . . . - .. ~   .
. . . - ..
                                          ..
~
                                              .
.
                                                        .
..
                                                            . _ . . . . .
.
                                                                                                                ;
.
                                                                          .
. _ . . . . .
                                                                                                .
.
                                                                                                    .
.
.
;


    .
.
  .
.
                                                  43
43
              -
OPLP-01. Revision 6. Fire Protection Program Document
                      OPLP-01. Revision 6. Fire Protection Program Document
-
              -
-
                      0FLP-01.1. Revision 12. Fire Protection Commitment Document
0FLP-01.1. Revision 12. Fire Protection Commitment Document
              -
OPLP-01.2 Revision 10. Fire Protection System Operability.
                      OPLP-01.2 Revision 10. Fire Protection System Operability.
-
                      Action, and Surveillance Requirements
Action, and Surveillance Requirements
              -
-
                      FPP 005. Revision 15. Fire Watch Program
FPP 005. Revision 15. Fire Watch Program
              -
-
                      FPP-008. Revision 24. Fire Protection Weekly inspection
FPP-008. Revision 24. Fire Protection Weekly inspection
              -
FPP 013. Revision 25. Transient Fire Load Evaluation
                      FPP 013. Revision 25. Transient Fire Load Evaluation
-
              -
FPP 014. Revision 17. Control of Combustible. Transient Fire loads
                      FPP 014. Revision 17. Control of Combustible. Transient Fire loads
-
!
!
                      and Ignition Sources
and Ignition Sources
              Plant tours were also performed to assess procedure complianc.e.
Plant tours were also performed to assess procedure complianc.e.
        b.   Obji.ervations and Findinas
b.
              The listed procedures were issued to implement the facility's fire
Obji.ervations and Findinas
              protection program.     These procedures contained requirements for program
The listed procedures were issued to implement the facility's fire
              administration, controls over combust 1 oles arid ignition sources, fire
protection program.
              watch duties and training, and operability requirements for fire
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
i
              protection systems and features. The 3rocedures were well written and
protection systems and features.
              met the licensee's commitments to the 1RC.
The 3rocedures were well written and
              General plant walkdown inspections were perfoimed by the inspector to
met the licensee's commitments to the
              verify: acceptable housekeeping; compliance with the ]lant's fire
1RC.
              prevention procedures such as control of transient com)ustibles:
General plant walkdown inspections were perfoimed by the inspector to
              operability of the fire detection and suppression systems: emergency               '
verify:
              lighting: and installation and operability of fire barriers, fire stop
acceptable housekeeping; compliance with the ]lant's fire
              and penetration seals (fire doors, dampers, electrical penetration
prevention procedures such as control of transient com)ustibles:
              seals, etc.),
operability of the fire detection and suppression systems: emergency
      c.     Conclusions
'
              General housekeeping was satisfactory. Fire retardant plastic sheeting
lighting: and installation and operability of fire barriers, fire stop
              and film materials were being used. Lubricants and oils were properly
and penetration seals (fire doors, dampers, electrical penetration
                stored in approved safety containers. Controls for combustible gas bulk
seals, etc.),
                storage and cutting and welding operations were being enforced.
c.
              Controls were being properly maintained for limiting t' alsient
Conclusions
                combustibles in designated separation zones and oth' restricted plant
General housekeeping was satisfactory.
            . areas.
Fire retardant plastic sheeting
      F5       Fire Protection Staff Training and Qualification
and film materials were being used.
      F5.1 EireBrioade
Lubricants and oils were properly
        a.     Insoection Stone (64704)
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
m
                                                                                                  44
1
                        The inspector reviewed the fire brigade organization and training
.
                        program for compliance with the NRC guidelines and program requirements.
1
                                                                                                                                                '
44
                b.     Observations and Findinos
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
The organization and training requirements for the plant fire brigade
                        were established by Fire Protection Procedure 0FPP-051. Loss Prevention
were established by Fire Protection Procedure 0FPP-051. Loss Prevention
                        Emergency Response 0ualification/ Training and Drill Program. The fire
Emergency Response 0ualification/ Training and Drill Program.
                        brigade for each of five shifts was composed of an operations support
The fire
                        fire protection technician shift incident commander (fire brigade
brigade for each of five shifts was composed of an operations support
                        leader) and at least four additional brigade members consisting of
fire protection technician shift incident commander (fire brigade
                        Auxiliary Operators. Chemistry Technicians and Maintenance personnel.
leader) and at least four additional brigade members consisting of
                        Each operations shift also had a Senior Reactor Operator / Reactor                                                       :
Auxiliary Operators. Chemistry Technicians and Maintenance personnel.
                        Operator Fire Brigade Advisor assigned to respond tr ' ires with the fire
Each operations shift also had a Senior Reactor Operator / Reactor
                        brigade.
:
                        As of the date of this inspection, there were a total of 48 fire brigade
Operator Fire Brigade Advisor assigned to respond tr ' ires with the fire
                        members 26 from operations and 22 from E&RC and Maintenance on the
brigade.
                        pic t fire brigade. The inspector verified that sufficient shift
As of the date of this inspection, there were a total of 48 fire brigade
                        personal were available to staff each shift's fire brigade with at
members 26 from operations and 22 from E&RC and Maintenance on the
                        least five qualified fire brigade members.
pic t fire brigade.
                        A review of the training records for the fire brigade members indicated
The inspector verified that sufficient shift
                        that the training, drill, respiratory and physical examination
personal were available to staff each shift's fire brigade with at
                        requirements for each active member were up to date and met the
least five qualified fire brigade members.
                        established site training requirements.
A review of the training records for the fire brigade members indicated
                        Fire Briaade Ecuioment:
that the training, drill, respiratory and physical examination
                        The fire brigade turnout gear and a fire response vehicle and trailer
requirements for each active member were up to date and met the
                        with fire brigade equi) ment was stored in the Operations / Fire Protection
established site training requirements.
                        equipment building. T1e_ inspector's inventory of the fire brigade
Fire Briaade Ecuioment:
                        equipment indicated that a sufficient number of turnout gear, consisting
The fire brigade turnout gear and a fire response vehicle and trailer
                        of coats, pants, boots, helmets, etc. , was provided to equip the fire
with fire brigade equi) ment was stored in the Operations / Fire Protection
                        brigade members expected to respond in the event of a fire or other
equipment building. T1e_ inspector's inventory of the fire brigade
                        emergency. The fire brigade turnout i., ear and fire fighting equipment
equipment indicated that a sufficient number of turnout gear, consisting
                        were being properly maintained,
of coats, pants, boots, helmets, etc. , was provided to equip the fire
              c.       Conclusions
brigade members expected to respond in the event of a fire or other
                        The fire brigade organization and qualification training met the
emergency.
                      -requirements of the site procedu.m                                             . Fire brigade turnout gear and fire
The fire brigade turnout i., ear and fire fighting equipment
                        fighting eouipment were being properly maintained.
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
l
e                                                                                   l
e
                                          45                                       j
l
  F6   Fire Protection Organization and Administration
45
  F6.1 Fire Protection Mananement and OraanizatioD
j
    a. Inspection Scope (64704),
F6
        The licensee's management and administration of the facility's fire
Fire Protection Organization and Administration
        protection program were reviewed for compliance with the commitments to
F6.1 Fire Protection Mananement and OraanizatioD
        the NRC and to current NRC guidelines.
a.
    b. Observations and Findinos
Inspection Scope (64704),
        During this report period the licensee reassigned the responsibility ior
The licensee's management and administration of the facility's fire
        the administration and implementation of the fire protection program
protection program were reviewed for compliance with the commitments to
        from the previous Loss Prevention Unit (LPU) to the Operations Shift and
the NRC and to current NRC guidelines.
        Support organizations. The LPU organization was dissolved.
b.
        The designated onsite manager responsible for the administration and
Observations and Findinos
        implementation of the fire protection program was the Operations
During this report period the licensee reassigned the responsibility ior
        Manager, This responsibility had been delegated to the Operations
the administration and implementation of the fire protection program
        Support Superintendent. The Operations Support Superintendent was
from the previous Loss Prevention Unit (LPU) to the Operations Shift and
        responsible for the station fire protection program, fire protection
Support organizations.
        surveillance testing of fire protection systems and equipment, and
The LPU organization was dissolved.
        ensuring that the aopropriate fire prevention procedures and fire
The designated onsite manager responsible for the administration and
        b:'igade programs were implemented. A Fire Protection Program
implementation of the fire protection program was the Operations
        C0ordinator reported to the Operations Support Superintendent.
Manager,
        Maintenarice of the 31 ant fire protection equipment was performed by the
This responsibility had been delegated to the Operations
        Maintenance Unit. cire protection related training was planned and
Support Superintendent.
        conducted by the Brunswick Training Se: tion. Coordination of the
The Operations Support Superintendent was
        station's fire protection program commitments and engineering functions
responsible for the station fire protection program, fire protection
        was provided by a fire protection system engineer in the Brunswick
surveillance testing of fire protection systems and equipment, and
        Engineering Support Section,
ensuring that the aopropriate fire prevention procedures and fire
    c. Conclusions
b:'igade programs were implemented.
        The coordination and oversight of the facility's fire protection program
A Fire Protection Program
        had been reassigned from the previous LPU organization to Shift
C0ordinator reported to the Operations Support Superintendent.
        Operations. The new organizational structure met NRC guidelines and the
Maintenarice of the 31 ant fire protection equipment was performed by the
        licensee's fire protection program requirements.
Maintenance Unit.
  F7   Quality Assurance in Fire Protection Activities
cire protection related training was planned and
  F7.1 Fire Protection Audits
conducted by the Brunswick Training Se: tion.
    a. Insoection Scope (64704)
Coordination of the
        The following audit report and the plant response to the issues were
station's fire protection program commitments and engineering functions
        reviewed:
was provided by a fire protection system engineer in the Brunswick
        -      Nuclear Assessment Section (NAS) Report B-FP-97-01. Brunswick Fire
Engineering Support Section,
              Protection and Loss Prevention Unit Assessment, dated
c.
              August 1. 1997.
Conclusions
                                                                                  i
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
46
        b. Observations and Findinas
b.
            The licensee's Nuclear Assessment Section performed an assessment of the
Observations and Findinas
            fire protection program and LPU on June 16-27. 1997. The report for
The licensee's Nuclear Assessment Section performed an assessment of the
            this assessment was Re) ort No. B FP-97 01. The assessment team
fire protection program and LPU on June 16-27. 1997.
            determined that the LPJ fire prevention and fire response activities
The report for
            were adequate; however, its implementation of the fire protection
this assessment was Re) ort No. B FP-97 01.
            )rogram was ineffective based on a number of program elements found to
The assessment team
            )e below acceptable standards. Findings from these assessments were
determined that the LPJ fire prevention and fire response activities
            categorized as strengths, issues, or weaknesses. The assessment report
were adequate; however, its implementation of the fire protection
            identified six program element issues and one weakness.
)rogram was ineffective based on a number of program elements found to
            The inspector reviewed the final audit report, the licensee's response
)e below acceptable standards.
            to the identified issues. the planned corrective actions, and the NAS
Findings from these assessments were
            evaluation of the response adequacy.
categorized as strengths, issues, or weaknesses.
            This NAS assessment of the facility's fire protection program was
The assessment report
            comprehensive and effective in identifying fire protection program
identified six program element issues and one weakness.
            performance deficiencies to management. The audit team identified
The inspector reviewed the final audit report, the licensee's response
            deficiencies in LPU'c management oversight of fire protection
to the identified issues. the planned corrective actions, and the NAS
            procedures, training, problem identification, procedure performance
evaluation of the response adequacy.
            standards, corrective actions, and personriel safety. Corrective actions
This NAS assessment of the facility's fire protection program was
            in response to the identified issues were substantial and included a
comprehensive and effective in identifying fire protection program
            fire protection reorganization to integrate the former LPU organization
performance deficiencies to management.
            into the shift Operations and Operations Sup) ort organizations under
The audit team identified
            direct management of the Operations Support Manager,
deficiencies in LPU'c management oversight of fire protection
        c. Conclusions
procedures, training, problem identification, procedure performance
            The 1997 Nuclear Assessment Section assessment of tite facility's fire
standards, corrective actions, and personriel safety.
            protection program was comprehensive and was effective in identifying
Corrective actions
            fire protection program performance deficiencies to management. Planned
in response to the identified issues were substantial and included a
            corrective actions in response to the audit issues were substantial and
fire protection reorganization to integrate the former LPU organization
            included a fire protection reorganization.
into the shift Operations and Operations Sup) ort organizations under
                                  V.   Manaaetment Meetinas
direct management of the Operations Support Manager,
      XI   Exit Meeting Summary
c.
            The inspector presented the inspection results to members of licensee
Conclusions
            management at tN conclusion of the ins)ection on January 8,1998. Post
The 1997 Nuclear Assessment Section assessment of tite facility's fire
            inspection briefings were conducted on )ecember 12, 1997. The licensee
protection program was comprehensive and was effective in identifying
            acknowledged the findings presented. The licensee stated that they had
fire protection program performance deficiencies to management.
            not determined if clearance records were required QA records.
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
A


                    .. __ _ _ _ _ _             _ _ _
..
                                                          - _ _ - _ - _ - _ _ - _ _ _ - - _
__ _ _ _ _ _
                                                                                              .
_ _ _
  .
- _ _ - _ - _ - _ _ - _ _ _ - - _
.
.
                                                        47
.
                                      PARTIAL LIST OF PERSONS CONTACTED
.
    Licensee
47
    A. Brittain. Manager Security
PARTIAL LIST OF PERSONS CONTACTED
    M. Christinziano, Manager Environmental and Radit lon Control
Licensee
                                                                                            _
A. Brittain. Manager Security
    W. Dorman. Supervisor Licensing and Regulatory Programs
M. Christinziano, Manager Environmental and Radit lon Control
    N. Gannon. Manager Maintenance
_
    J. Gawron. Manager Nuclear Assessment Section
W. Dorman. Supervisor Licensing and Regulatory Programs
    S. Hinnant. Vice President. Brunswick Steam Electric Plant
N. Gannon. Manager Maintenance
    K. Jury. Manager Regulatory Affairs
J. Gawron. Manager Nuclear Assessment Section
    R. Krich, Chief Engineer. Nuclear Engineering Department
S. Hinnant. Vice President. Brunswick Steam Electric Plant
    B. Lindgren. Manager Site Su) port Services
K. Jury. Manager Regulatory Affairs
    J. Lyash. Manager Brunswick Engineering Support Section
R. Krich, Chief Engineer. Nuclear Engineering Department
    R. Mullis. Manager Operations
B. Lindgren. Manager Site Su) port Services
    Other licensee employees or contractors included office, operation,
J. Lyash. Manager Brunswick Engineering Support Section
    maintenance. chemistry, radiation, and corporate personnel.
R. Mullis. Manager Operations
                                        _
Other licensee employees or contractors included office, operation,
maintenance. chemistry, radiation, and corporate personnel.
_


.
.
                                          48
48
                            INSPECTION PROCEDURES USED
INSPECTION PROCEDURES USED
  IP 37550:   Engineering
IP 37550:
  IP 37551:   Onsite Eng11eering
Engineering
  IP 61726.   Surveillance Observations
IP 37551:
  IP 62700:   Maintenance Program implementation
Onsite Eng11eering
  IP 62707:   Maintenance Observations
IP 61726.
  IP 64704:   Fire Protection
Surveillance Observations
  IP 71707:   Plant 0)erations
IP 62700:
  IP 71714:   Freeze )rotection
Maintenance Program implementation
  IP 71750:   Plant Support Activities
IP 62707:
  IP 92700:   Onsite Followup of Written Reports of Nonroutine Events at Power
Maintenance Observations
              Reactor Facilities
IP 64704:
  IP 92901:   Followup - Operations
Fire Protection
  IP 92902:   Followup - Maintenance
IP 71707:
  IP 92903:   Followup - Engineering
Plant 0)erations
                        ITEMS OPENED, CLOSED, AND DISCUSSED
IP 71714:
  Opened
Freeze )rotection
  50-325(324)/97-13-01     VIO   Failure to Retain TS Required QA Record (Section
IP 71750:
                                07.2)
Plant Support Activities
  50 325(324)/97 13-02     VIO   Inadequate Procedure for the Conduct of E0
IP 92700:
                                Preventive Maintenance (Section M1.3, El.4.b.6)
Onsite Followup of Written Reports of Nonroutine Events at Power
  50 325/97-13-03         VIO   Failure to Note Abnormal TS Surveillance Values
Reactor Facilities
                                (Section M3.1)
IP 92901:
  50 325(324)/97-13-04     IFl   Review of Licensee Records and Engineering
Followup - Operations
                                Evaluations to Establish the Fire Resistant
IP 92902:
                                Capabilities of Fire Rated Silicone foam
Followup - Maintenance
                                Penetration Seals (Section F2.2)
IP 92903:
  50-325(324)/97-13-05     URI   UFSAR Discrepancy Fire Doors (Section F2.4)
Followup - Engineering
  50 325(324)/97-13 06     IFl   Revisions to Procedure EGR-NGGC-0153 (Section
ITEMS OPENED, CLOSED, AND DISCUSSED
                                El.1)
Opened
  50-325(324)/97-13-07     IFl   Review Technical Evaluation of Terminal Block
50-325(324)/97-13-01
                                Current Leakayc and the Effect on EQ Equipment.
VIO
                                (Section El.4.b.4)
Failure to Retain TS Required QA Record (Section
  Closed
07.2)
  50-325/96-15-01         URI   Vessel Disassembly Without Secondary Containment
50 325(324)/97 13-02
                                (Section 08.1)
VIO
  50-325(324)/97 02-01     V10   Locked Valve Out of Position (Section 08.2)
Inadequate Procedure for the Conduct of E0
  50-325/97 12 03         URI   Recirculation Pump Run back (Section 08.3)
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
49
                          50-325(324)97-12-04   URI   Diesel Generator Low Voltage Auto Start Defeated
50-325(324)97-12-04
                                                      (Section 08.4)
URI
                          50 325(324)/96-017-00 LER   Invalid Loss of Coolant Accident (Section M8.1)
Diesel Generator Low Voltage Auto Start Defeated
                          50_-325/97_009-00     LER   Missed Increased Frequency inservice Testing
(Section 08.4)
                                                      Requirement (Section M8.2)
50 325(324)/96-017-00
                          50-325/97-001-00     LER   Rod Block Monitor Surveillance inadequacy
LER
                                                      (Section M8.3)
Invalid Loss of Coolant Accident (Section M8.1)
                          50-325(324)/95-022 00 LER   High Pressure Coolant injection System Discharge
50_-325/97_009-00
                                                      Flow Element Gasket Leak (Section M8.4)
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
50 325/97-05-02
                                                      Abnormal   CS Sp)arger Break Detector Indication
IFl
                                                      (Section Md.5
Abnormal CS Sp)arger Break Detector Indication
                          50 325/97-05-03       VIO   Inadequate CS Surveillance Procedure (Section
(Section Md.5
                                                      M8.5)
50 325/97-05-03
                          50 325/97-02         LER   Core Spray Header Differential Pressure
VIO
                                                      Instrumentation Inoperable (Section M8.5)
Inadequate CS Surveillance Procedure (Section
                          50-325(324)/97-02-04 VIO   Failure to implement Requirements of the
M8.5)
                                                      Maintenance Rule (Section M8.6)
50 325/97-02
                          50-325(324)/97-08-04 URI   Control of EDBS Information (Section E8.1)
LER
                          50-325(324)/97-04     LER   Spent Fuel Shipping Cask Handling Activities
Core Spray Header Differential Pressure
                                                      (Section E8.2)
Instrumentation Inoperable (Section M8.5)
                          50-325(324)/96-14-05 IFI   Effect of EQ Accuracy on Instrument Setpoint
50-325(324)/97-02-04
                                                      Calculations (Section E8.3)
VIO
                          50-325(324)/97-02-08 VIO   Failure to Implement an Inspection Program for
Failure to implement Requirements of the
                                                      Safety-Related Miscellaneous Structural Steel
Maintenance Rule (Section M8.6)
                                                      (Section E8.4)
50-325(324)/97-08-04
                          50-325(324)/97-08 07 VIO   Failure to Initiate Condition Reports to
URI
                                                      Document Nonconforming EQ ltems (Section E8.5)
Control of EDBS Information (Section E8.1)
  r
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
.
-
.
}
}}
}}

Latest revision as of 03:06, 24 May 2025

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
ML20199G750
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 01/23/1998
From:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20199G672 List:
References
50-324-97-13, 50-325-97-13, NUDOCS 9802040338
Download: ML20199G750 (50)
v • d • e


See also: IR 05000324/1997013

Text

_ _ _ _ _ _ _ _ _ _ _ _

.

.

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

ADOCK 05000324

G

PDR

Enclosure 2

_ _ _ _ _ _ _ _ _ _ _ .

i

\\

.

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

The inspector concluded that u.e cold weather program has been

e

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),

The inspector concluded that the Plant Nuclear Safety Committee meeting

e

provided an effective review of Unit I readiness for restart (Section

07.1).

Inspe.; tor review determined that clearance records were not retained in

e

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

2

Maintenance

Movement of the spent fuel shi) ping cask was perforrxo in accordance

e

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

.

accomplished in a timely manner.

The maintenance and material condition

i

!

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-

1974Property "ANSI code" (as page type) with input value "ANSI N45.2.9-</br></br>1974" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The electrical

..

bus was not restored, in this case, for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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

l

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)

off-9as radiation monitors on aage 26 were between 1570 and 1780

i

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

14

s

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

Themfore. the licensee could not evaluate whether the alarm setpoint

I

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

-

,

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.

.

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

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.

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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

. . .

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o

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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.

.

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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

.

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_ _________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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.

>

+

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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

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.

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) period prior to initiation of CR 97-

The inspectors reviewed 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

by E0P-03 SCCP Secondary Containment Control Proccdure (SCCP),

e

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

CR 97-02015.ystem) and were familiar with the problem addrc ses by

protection s

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

  1. 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

gathering process.

For other personnel, the training included

s

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 reshonded to this violation in a letter dated

September 2. 199

Subject: Reply to Notice of Violation.

The

.

,

_ _ ,

_ . - . . _ . _ ,

.

. . . _ _

.

c

- , . -

_

-

-_

__,

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

3

1

Fire Suppression System

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

-, ,, - . . .

-. - -

. .

. - - - . - . . .

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.

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- .

~ -

. -

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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

.

thermal mass of penetrating items

e

clearances of penetrating items

e

unexposed surface temperatures

e

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.

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_ _ _ - - - _ - - . - - - . - - -

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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

licensee stated that industry documentation is available to support

i

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.

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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:

>

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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)

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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.

__

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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

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.

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.

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)

)

_ _ _ _ _ _ _ _ _ _

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.

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)

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