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{{Adams
#REDIRECT [[IR 05000267/1988003]]
| number = ML20150F008
| issue date = 03/25/1988
| title = Insp Rept 50-267/88-03 on 880201-29.Violations Noted.Major Areas Inspected:Licensee Action on Previously Identified Findings,Operational Safety Verification,Followup of Unusual Events,Esf Walkdown & Physical Security Observation
| author name = Farrell R, Michaud P, Westerman T
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| addressee name =
| addressee affiliation =
| docket = 05000267
| license number =
| contact person =
| document report number = 50-267-88-03, 50-267-88-3, NUDOCS 8804040236
| package number = ML20150F000
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 12
}}
See also: [[see also::IR 05000267/1988003]]
 
=Text=
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                                                                          APPENDIX B
4
                ,                                          U. S. NUCLEAR RECULATORY COMMISSION
                                                                                                                                '
              >,
                                                                          REGION IV                              ,
      1
                                                                                                                                '
                                  NRC Inspection Report:      50-267/88-03                  License:  DPR-34          -
        .
                                  Docket:  50-267.
                                  Licensee:  Public Service Company of Colorado (PSC)
                                  Facility Name:    Fort St. Vrain Nuclear Generating Station
                                                    Fort St. Vrain (FSV) Nuclear Generating Station, Platteville.
                                                                      ~
                                  Inspection At:                                                                                .
                                                        Colorado                                                                I
                                                                                                                                t
                                  Inspection Conducted:      February 1-29, 1988                                                ;
                                                                        /f                            )~8O8
                                                                                                                                i
                                  Inspectors:              [
                                                  R. E. Farrell,"Senior Resident Inspector (SRI)      Date
'
                                                    Y(A/r
                                                  F. W. Michaud, Rbsident Inspector (RI)
                                                                                                      3*/79
                                                                                                      Date
                                                                                                                          ,
                                                                                                                                !
                                  Approved:      }          tt)    e d                                3/2b
                                                                                                      D6te'
                                                  T.'F. Westerman, Chief
                                                    Reactor Projects Section B
    '
;
!
,
W
W
  ,
t '
                                                                                                                            i
            8804040236 890331
            PDR              ADOCK 05000267
            0                              DCD
 
  _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _
              . .                      .
                            .        .
                                                                                  2
                                          Inspection Summary
                                          Inspection Conducted February 1-29,1988 (Report 50-267/88-03)
                                          Areas Inspected:  Routine, unannounced inspection of folicwup of licensee
                                          action on previously identified findi;.gs, operational safety verification,
                                          followup of unusual event, engineered safety features walkdown, monthly
                                          surveillance observation, monthly maintenance observation, radiological
                                          protection, and physical security observation.
                                          Results: Within the eight areas inspected, one violation was identified (the
                                          failure to implement and follow procedures for maintenance and operations
                                          activities, paragraph 4).
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I
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  ._--____ _ ___ _ _ _ _ _ _ - _ _ _ _ _ _ - - - _ _ _ _ _ _ - _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ - _ _ _ _ _ _ _ _ _                                                          _
  '.              .
                                                '
                                                              .
            .                                      .
                                                                                                                                                                                                                                            3
                                                                                                                                                                                                                                  DETAILS
                                                                                1.                                    Persons Contacted
                                                                                                                      FSV
                                                                                                              *L. Brey, Manager, Nuclear Licensing and Fuels
                                                                                                              "M. Ferris, Manager, Quality Assurance (QA) Operations
                                                                                                              *C. Fuller, Manager, Nuclear Production
                                                                                                              *M. Holmes, Manager, Nuclear Licensing
;
                                                                                                              *F. Novachek, Manager. Technical / Administrative Services
                                                                                                              *P. Tomlinson, Manager, QA                                                                                                                ,
                                                                                                              *D. Warembourg, Manager, Nuclear Engineering
                                                                                                              *R. Williams Jr. , Vice President, Nuclear Operations
                                                                                                              *J. Reesy, Staff Assistant, Nuclear Engineering
,                                                                                                            *F. Borst, Nuclear Training Manager
                                                                                                              *M. Deniston, Shift Supervisor
                                                                                                              *S. Hofsetter, Nuclear Licensing
                                                                                                              *M. Block, Superirtendent, Nuclear Betterment                                                                                              l
                                                                                                              *L. Scott, Manager, QA Service
                                                                                                              *R. Sargent, Assistant to Vice President, Nuclear Operai1ons
                                                                                                              *R. Webb, Maintenance Supervisor
,
                                                                                                                    The NRC inspectors also contacted other licensee and contractor personnel
                                                                                                                    during the inspection.
                                                                                                                      * Denotes those attending the exit interview conducted March 8, 1988.
!
                                                                              2.                                    Followup of Licensee Action on Previously Identified Findings
                                                                                                                      (Closed) Open Item 267/8507-06: Shorten Time Between Change Notice (CN)
                                                                                                                      Issue And Notation On Drawing - In some cases, a caution that changes had
.                                                                                                                    been made under a CN was not reflected on the affected drawings for 30
                                                                                                                    days or more af ter a CN was issued. This presented a concern that a
                                                                                                                    modified system or component could be in service for that amount of time
                                                                                                                    without adequate drawings. By utilizing a computerized document update
                                                                                                                      information system, the licensee has shortened the time involved to mark
                                                                                                                    all affected drawings to approximately 1 week, with the drawings in the
2                                                                                                                    control room, shif t supervisor's of fice, and records center updated the
                                                                                                                      same day a CN issue notification is received. The NRC inspector verified
'
                                                                                                                      these activities are taking place by direct observations and a review of
;                                                                                                                    documentation. This item is closed.
                                                                                                                      (Closed) Open Item 267/8507-07: Devcon Epoxy Only Qualified to 200 F.
                                                                                                                      Epoxy used to attach thermocouples to control rod drive assemblies was
                                                                                                                      qualified to enly 200*F, while actual operating temperatures can exceed
                                                                                                                      200 F. Two tests were performed by the licensee to establish this
                                                                                                                      adhesive's acceptability. One test performed under Fuel Handling
                                                                                                                      Procedure 100-31 involved a visual examination and measurement of force
,
  ,e, , ,                    e                                        +,,_mmr.n.,rs.                                                                                  -r,      ,-g      ~ a,---., -----.,--w.m..---,,, e,-e s---,ww,w<-  - - , -
 
                                          __
      *
  ..      .
    .  .
                                                    4
              required to remove the epoxy from a CRD element, which had been subjected
l              to varying power operating conditions in the reactor core between 1979 and
l              1984. The second test, T-288, involved subjecting epoxy to greater than
1              300 F temperature and then performing a pull test to verify that
              thermocouples remained sufficiantly attached. Based on these tests, the
,
              licensee concluded the Devcon ep;xy was acceptable for use in applications
'
              up to 300 F. The NRC resident inspectors reviewed the licensee's tests
              and evaluation and found them acceptable.      This item is closed.
            3. Operational Safety Verification
              The NRC resident inspectors reviewed licensee activities to ascertain that
              the facility is being operated safely and in conformance with regulatory
              requirements and that the licensee's management control system is
              ef fectively discharging its responsibilities for continued safe operation.
              The NRC resident inspectors toured the control room on a daily basis
              during normal working hours and at least weekly during backshif t hours.
              The reactor operator and shif t supervisor logs and Technical Specification
              compliance logs were reviewed daily. The NRC resident inspectors observed
              proper control room staffing at all times and verified operators were
              attentive and adhered to approved procedures. Control room
                instrumentation was observed by the NRC inspectors and the operability of
              the plant protective system and nuclear instrumentation system were
              verified by the NRC resident inspectors on each control room tour.
              Operator awareness and understanding of abnormal or alarm conditions were
              also verified.    The NRC resident inspectors revieved the operations order
              book, operations deviution report (0DR) log, clearance log, and temporary
              configuration report (TCR) log to note any out-of-service safety-related
                systems and to verify compliance with Technical Specification
                requirements.
              The licensee's station manager and superintendent of operations were
              observed in the control room on a daily basis, with the superintendent of
              operations frequently in the control room during the day and during
                special tests or evolutions.
                The NRC resident inspectors verified the operability of a safety-related
                system on a weekly basis.    The PCRV overpressure protection system,
                120 VAC vital power distribution system, reactor plant cooling water
                system, and firewater system were verified operable by the NRC resident
                inspectors during this report period.      During plant tours, particular
                attention was paid to components of these systems to verify valve
                positions, power supplies, and inst umentation were correct for current
                plant conditions.    General plant condition and housekeeping were
                acceptable.
                Shift turnovers were observed at least weekly by the NRC resident
                inspectors. The information flow appeared to be good, with the shift
                                                                                          l
                                                                                          l
                                                                                          l
                                                                                          ;
 
                        -        . - .  .-      .
  '
..    ..
.  .
                                                      5
            supervisors routinely soliciting comments' or concerns from reactor
            operators, equipment operators, and auxiliary tenders.
            No violations or deviations were identified in the review of this program
            area.
          4. Followup of Unusual Event
            On February 10, 1988, at 3:47 p.m. (MST), "A" helium circulator tripped
            due to a low speed signal with the reactor at 75 percent power. The
            circulator trip resulted in a reactor runback to between 50 percent to
            60 percent reactor power and then reactor power was further reduced by the
            plant operators to 25 percent power. While attempting to balance
            feedwater between Loop 1 and Loop 2, an upset in the helium circulator
            auxiliaries supplied by feedwater resulted in the tripping of "B"      and "D"
            helium circuiators at 4:07 p.m. (MST).
            The tripping of two circulators (A & B) in one loop r.;ulted in a loop
            shutdown (ESF actuation).        The reactor operators manually scrammed the
            reactor from 25 percent power with only one helium circulator running.
            At 6:40 p.m. (MST), the licensee identified that an unplanned release was
            occurring and an unusual event was declared. An operator had been
            dispatched to vent the surge tank associated with the liner cooling water
            system. The licensed operator dispatched to perform this function
            inadvertently opened the wrong valve venting the tank to the plant stack
            rather than to the gaseous radwaste system. The total release over
            approximately 200 minutes was small. (4.26 X 105 microcuries of noble gas
            activity)
            The plant maintained forced circulation cooling at all timas.        The SRI
            responded to the event and was onsite all night. The Colorado Department
            of Health was in contact with the site and was briefed by the licensee as
            well as the SRI.
            The licensee has subsequently determined that the "A" helium circulator
            trip occurred due to an apparent interchange of speed indication signal
            cables during a recent equipment calibration. The trip occurred when the
            "B" helium circulator was placed in manual control for calibration.
            a.    Background
                  The unusual event of February 10, 1988, and associated unplanned
                  release started with the trip of helium circulator "A".
                  Helium circulator speed cable daily calibration was in process when
                  circulator "A"    tripped.
                  When a circulator's speed cables are calibrated, the circuiator is
                  taken from auto to manual control to minimize the chances of a trip.
                                                                                .
 
        _ _ _ _ _ _ .
  .
      .
    .                .
                                                                      6
                                        Helium circulator "A" speed cables had been successfully calibrated
                                        and circulator "A" returned to auto control. Helium circulator "B"
                                        was placed in manual control and calibration of the "B" circulator
                                        speed cables was in process when circulator "A" tripped.
                                        The licensee determined that on February 2, 1988, while calibrating
                                        the speed modules (SM) on circulator "A", SM 2109 could not be
                                        balanced while getting its signal from cable 18194.    The technician
                                        decided to check if the problem was in SM-2109 or in the cable 18194.
                                        The licensee suspected the speed problems were in the cables. Seven
                                        spare speed cables are available from each circulator's SM. The
                                        technician unplugged cable 18194 from SM 2109 and plugged in
                                        cable 18133. With cable 18133 installed, SM-2109 balanced and was
                                        left in this configuration by the technician.    Cable 18133 does not
                                        sense circulator "A" speed but is a spare speed cable from the "B"
                                        circulator,
                          b.          Design Information
                                        There are two speed indications from each circulator:    a steam
                                        turbine speed indication and a water turbine speed indication. The
                                        water turbine speed indicator is much easier to read than the steam
                                        indicator and generally the one the operators use.    Since both drives
                                        are on a common shaft, the speed should be the same regardless of
                                        which turbine is driving the circulator.
                                        There are 12 speed cables coming from the speed modules of each
                                        helium circulator. Four of these cables are utilized for speed
                                        control. One cable for steam turbine speed, one cable for water
                                        turbine speed, and two spares.
                                        Eight cables from each circulator are dedicated to the plant
                                        protection system (PPS).    Three of these cables are used at one time
                                        (one for each logic channel). Five cables are dedicated spares.
                          c.          Speed Control
                                        The speed control circuitry looks at the water turbine indicated
                                        speed and the steam turbine indicated speed and controls from the
                                        higher of the two indicated speeds (no difference if everything
                                        working correctly).
                                        As long as the "B" circulator speed was less than or equal to the "A"
,
                                        circulator speed, the control system saw no problem and chose the "A"
!
                                        circulator steam turbine speed to control circulator "A" With
l                                      cable 18133 (a "B" circulator speed cable) controlling SM-2109 (the
                                        "A" circulator water turbine SM) the problem arose during calibration
                                        of "B" circulator speed when the "A" circulator was in auto control
                                        and the "B" circulator speed exceeded the "A" circulator speed,
l
1
                        --    - _ - _ .  . - .  .-.
 
                                      .
                                                          .      .
                                                                                _-      _ _ _ _ __
                        .
        .  '.
      .    .
                                                  7
                  .When this happened, the control circuit for "A" circulator, selecting
                  the higher speed indication, selected the "A" circulator water
                  turbine speed. -This was actually the "B" circulator speed, since a
                  "B" cable was feeding this speed module. This falsely told the
                  control circuit that the "A" circulator was running faster than the
                  control circuit required, so the control circuit began closing the
                  "A" circulator steam speed valve.
                  Since the control circuit was actually reading "B" circulator speed
                  it saw no change in the "A" circulator speed indication and continued
                  to close down the "A" circulator speed valve. When the "A"
                  circulator reached the low setpoint of the circulator
                  speed-to-feedwater flow program, which forces a limit on primary to
                  secondary flow ratio, the the PPS which was correctly reading
                  circulator "A" speed tripped the circulator.
              d.  Findings
                  The technician calibrating the SM was utilizing licensee
                  Procedure SR-RE-17-W, Issue 10, "Circulator Speed Modifier Weekly
                  Check."
                  The procedure did not address cable termination.
                  When the technician removed the installed cable (18133) he was no
                    longer performing surveillance activities, but was performing
                  maintenance activities. Maintenance activities are governed by the
                    licensee's Administrative Prosedure P-7, Issue 12, "Station Service
                  Request Processing." Procedure P-7 as modified by Procedure
                  Deviation Request 88-0006, dated January 13, 1988, specifically
                    states, in Section 2.0, that the procedure applies to corrective and
                    preventative maintenance and not to calibration activities.
                    Procedure P-7 is the licensee's procedure for controlling maintenance
                    activities.    Procedure P-7 requires initiation of a Station Service
                    Request to authorize, document, and control maintenance activities.
                    Failure to follow Procedure P-7 is an apparent violation of NRC
                    regulations (267/8803-01).
                    The operator venting reactor plant cooling water system surge tanks
                    was guided by System Operating procedure (SOP) 46, Issue 39, "Reactor
                    Plant Cooling Water System." SOP-46 in Step 3.7, "Venting the Vapor
                    Space in T-4601 or (T-4602)," details the steps for venting the
                    reactor plant cooling water surge tank vapor space to the gas waste
                    system. The steps call for first opening V-4653 for Surge
                    Tank T-4601 (V-4654 for Surge Tank T-4602). Then the operator is to
                    open V-461691 for Surge Tank T-4601 (V-461692 for Surge Tank T-4602).
                    Opening these two valves for each surge tank vents the vapor space of
                    each tank to a common line leading to the gas waste system. When
                    these steps are completed, the operator opens Valve V-46193, which
. . . . . .
 
                                                        -____      _ _ _ _ _ _ _ .
  . '.
.    .
                                              8
              opens the common line from the two surge tanks to the gas waste
                system relieving the pressure in the , urge tanks.
              All of the valves mentioned in the preceding paragraph are manual
                valves. Adjacent to the valves, V-461691 on Tank T-4601 and V-461692
                on Tank T-4602, are hand operated valves, V-461634P and k-461635P,
                respectively. Opening Valve V-461634P after opening Valve V-4653
                vents Surge Tank 1-4601 to the plant exhaust stack. Opening Valve
                V-461635P af ter opening Valve V-4654 vents Surge Tank T-4602 to the
                plant exhaust tank.
                The valves are now clearly marked as to function. At the time of the
                incident, the valves were marked with small stamped metal tags
                identifying the valves by number.
                Procedure 50P-46 in Step 3.7 clearly listed the valves to be opened.
                The valves were identified in the procedure by valve number
                corresponding to the valve numbers attached to the valves. The
                operator opened either or both Valves V-461634P and V-461635P, rather
                than V-461691 and V-461592. This vented the gaseous content of
                Tanks T-4601 and/o" T-4602 to the plant stack resulting in an
                unplanned radioactive release. The failure to follow
                Procedure 50P-46 is second example of Violation (267/8803-01).
      5. Engineered Safety Features (ESF) Walkdown
          The NRC resident inspectors performed a walkdown of all accessible
          portions of the prestressed concrete reactor vessel (PCRV) overpressure
          protection system to verify its operability.        Sections 4.3.6 and 6.8 of
          the FSAR and Technical Specifications 3.2, 3.3, 4.2.7, and 5.2.1 were
          reviewed by the NRC resident inspectors to ensure familiarity with the
          system and requirements. The as-found system configuration was compared
          with drawing PI-11-5 to check their agreement. Valve positions and
          labeling were verified to be correct by the NRC resident inspectors,
          including the installation of lotking devices on valves where required.
          All cortions of the system were physically inspected, w th the exception
          of the internals of the PCRV safety valve tank T-1101 which contains the
          relief valves and rupture discs. These components will be inspected
          during the next outage when T-1101 is opened. During this inspection,
          attention was paid to equipment conditions, housekeeping, and any items
          which could degrade performance. The overall condition of this system was
          considered good.
          No violations or deviations were identified in the review of this program
          area.
      6. Monthly Surveillance Observation
          The NRC resident inspectors observed the licensee's performance of
          selected surveillance activities as listed below. The surveillance
          procedures were reviewed for conformance with Technical Spe;ification
 
    .  .
  o  .
                                                  9
            requirements and to ensure they had been properly reviewed and approved
            prior to commencing any tests. The NRC resident inspectors witnessed
            portions of the preparations, conduct, and v/ stem restoration for each of
            these surveillance tests. Test results were independently reviewed by the
            NRC resident inspectors to ensure they met applicable Technical
            Specification requirements. Surveillance activities observed during this
            reporting period included:
                  SR 5.4.1.1.8.b-M, "Reheat Steam Temperature Scram Test," performed on
                  February 1,1988. This surveillance tests each hot reheat steam
                  temperature scram channel to verify alarms, actuations, and
                  indications.  The as-found values were measured and recorded,
                  acceptance values calculated and independently verified, and
                  calibration of the bystable amplifiers and thermocouple amplifiers
                  was checked at 600 F, 900 F, and 1200 F utilizing test signals.
                  These amplifiers were adjusted as required in accordance with this
                  procedure and the as-left values were recorded.    No discrepancies
                  were noted.
                  SR 5.10.8-M, "Monthly Check of Fire Hose Stations," performed on
                  February 2, 1988. This surveillance verified the condition of each
                  fire hose station in the reactor and turbine buildings, and was
                  independently versfied by the NRC resident inspectors.    Each
                  station's hose valve was verified shut and not leaking, hoses and
                  nozzles properly connected, and general equipment conditions
                  observed.  No discrepancies were noted.
                  E3R 8.1.lbc-M, "Radioactive Gaseous Effluent Systein Test," performed
                  on February 25, 1988. This surveillance test verifies the operation
                  of the gaseous waste release system automatic functions. Instruments
                  which provide inputs to cause automatic isolation and ventilation
                  system realignments were tripped using a test signal, then each
                  associated damper or valve which was repositioned by the automatic
                  signal was verified to be in its proper position.    The instruments
                  and equipment were then restored to their normal lineup. No
                  discrepancies were noted.
            No violations or deviations were identified in the review of this program
            area.
          7. Monthly Maintenance Observation
            On February 4,1988, the licensee noticed the pressure in the emergency
            feedwater supply to the Loop 1 helium circulator Pelton wheel drives was
            equal to the feedwater header pressure (approximately 3000 psia). This
:
            condition indicated a problem with Pressure Control Valve PV-21243, which
'
            should reduce the pressure to approximately 1700 psi. The licensee took
            the emergency feedwater header out of service at 5:57 a.m. (MST), on
            February 5, 1988, to perform repairs on PV-21243 and entered Technical
            Specification Limiting Condition for Operation (LCO) 4.0.3, since the
;
            conditions of LCO 4.3.4, "Emergency Condensate and Emergency Feedwater
i
i
i
 
                                                                  -      -                  _ _ _ _ _
          '
      .      .
    .  ..
                                                  10                                                  o
,
              Headers LCO," were no longer satistted-    LCO 4.0.3 requires the reactor to
              be ;hutdown in an orderly manner within a 24-hr Jr period. Also c' :icaale
              and providing a 24-hour grace period was LCO 4.2.2.a, "Operable C. culator              -
              LCO." Repairs were made to valve PV-21243, which included replacement of
              the valve trim. The associated pressure controller, PIC-21243, was
              calibrated in accordance with Procedure RP-EQ-16, Issue 2, dated
              October 15, 1986. The NRC resident inspectors observed the repairs and
              calibration, which were completed satisfactorily. No aiscrepancies were
              noted. The emergency feedwater header was retu ned to service at
              1 a.m. (MST), on February 6, 1988, and LCO 4.0.3 and 4.2.2.a were formally
              exited at 5:15 a.m. (MST), after allowing the system to run following its
              return to service.
              The NRC resident inspectors also followed the licensee's actions to
              correct the problems in the helium circu'ator speed cables. The
              circulator speed signals to both the indicators and the plant protective
              system had been exhibiting erratic behavior at the elevated temperatures
              associated with operation at higher power levels. Troubleshooting
              following the February 10, 1988, event, described in paragraph 4 of this
              report, indicated a problem with the twinax cable "Cannon" connectors at
              the helium circulators. These special connectors have the male end
              attached to the circulator housing and the female end attached to the
              cables. These female pin connectors have a spring-like device which in
              some cases had relaxed, allowing a slight gap in the pin connection at the
              elevated temperatures. The connectors on each of the four helium
              circulators were disassembled and both the nale ar.d fema'e pins were
              checked with a micrometer to ensure their size was within a tolerance of
              0.060 inch to 0.064 inch. A number of female pins were replaced, and the
              connectors reassembled. Since returning to power on February 12, 1988,
              the licensee has experienced no significant problems wish the helium
              circulator speed cables or the associated indications and protective
              circuitry.
  4
              At 10:40 p.m. (MST), on February 25, 1988, the license 4 experienced a
                turbine trip from approximately 50 percent power due to a <=lse low main
                steam pressure signal. On investigation, the licensee discovered the root
                valve to Main Steam Pressure Transmitter PT-5220 was nearly shut. This
                valve had been repacked on February 11, 1988, and was left in a nearly
                shut position following this work.    The valve was open enough to allow the
                main steam pressure to equalize across it before the turbine was placed in
                service. The valve's new paa,ing shifted, evidenced by the fact that the
                valve developed a packing leak about the time of the turbine trip, which
                allowed the pressure downstream of the ,alve to be relieved. This re6fced
                pressure was sensed by PT-5220, which then caused a turb'ne trip.
                The NRC resident inspectors found no instructions in Maintenance
                Procedure MP-2115 to return a valve to its as-found pcsttion following
                maintenance.    Although this is not safety-related equipment, the lack of a
                step to return the equipment to service following maintenance is of some
                concern. The licensee considers the potential probiers associated with
                this significant and will revise all m.iintenance precedures fo valves to
                                                                                                        j
                                                              _      .. _          _
 
          *
    .      .
  . . . . .
                                                    11
                record the as-found position before commencing maintenance and to return
                the valve to that position or leave it in another position with the shift
                supervisor's knowledge and consent following completion of the maintenance
                activity.  The NRC resident inspectors will monitor the licensee's
                implementation of these measures.
                No violations or deviations were identified in the review of this program
                area.
              8. Radiological protection
                The NRC resident inspectors observed the licensee's activities in this
                area to verify their conformance with policies, procedures, and regulatory
                requirements.
                Health physics professionals were observed on all shifts, performing plant
                tours, area surveys for radiation levels and radit. .ive contamination,
                and checking the operability of area radiation man, toes and continuous air
                samplers. The NRC resident inspectors verified tha the results of area
                surveys were posted at entrances to radiation areas and in other
                appropriate locations.    Health physics supervisors and personnel were
                aware of the plant status and activities which involved potential
                radiological concerns.
                The NRC resident inspectors observed that health physics personnel were
                present and available to provide astistance whenever workers are required
                to enter a radiologically controlled area.
                No violations or deviations were identified in the review of this program
                area.
              9. Physical Security Observation
                The NRC resident inspectors vcrified that there was a lead security
                officer (LS0) on duty authorized by the facility security plan to direct
;                security activities onsite for eac's shif t.  The LSO did not have duties
                that vould interfere with the direction of security activi+1es.
                The NRC resident inspectors verified, randomly and on the backshift, that
                the minimum number of armed guarcs required by the facility's security
                plan were present. Search equipment, including the X+ ray machine, metal
                detector, and explosive detector, were operational or a 100 percuat
                hands-on search was being utilized.
                The protected area barrier was surveyed by the NRC resident inspectors.
                The barrier was properly maintained and was not compromised by erosion,
                openings in tl.. fence fabric or walls, or proximity of vehicles, crates or
                other objects that could be used to scale the barrier.    The NRC resident
                inspectors observed tne vital area barriers were well maintained and not
                ccmpromised by obvious breaches or weaknesses. Th NRC resident
 
      ~
- .    .
...o
                                                12
              inspectors observed that persons granted access to the site were badged
              indicating whether they had unescorted or escorted access authorization.
              No violations or deviations were identified in the review of this program
              area.
          10. Exit Meeting
              An exit meeting was conducted on March 8, 1988, attended by those
              identified in paragraph 1. At this time, the NRC resident inspectors
              reviewed the scope and findings of the inspection.
                                .  -
                                        .                -.                      ..
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