IR 05000213/1988002

Safety Insp Repts 50-213/88-02 on 880113-0229.No Violations Noted.Major Areas Inspected:Outage Activities,Radiation Protection,Fire Protection,Security,Maint,Surveillance Testing & Events Occurring During Insp Period

ML20148A885
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	03/11/1988
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20148A878	List: IR 05000213/1988002 ML20148A876
References
50-213-88-02, 50-213-88-2, IEB-87-044, IEB-87-44, NUDOCS 8803210219
Download: ML20148A885 (12)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /88-02 ,

Docket N License N DPR-61 Licensee:

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Connecticut Yankee Atomic Power Company

.P. O. Box 270

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Hartford, CT 06101 Facility: Haddam Neck Plant, Haddam Neck, Connecticut-Inspection at: Haddam Neck Plant Inspection dates: January 13 - February 29, 1988 Inspectors: Andra A. Asars, Resident Inspector John T. Shediosky, Senior Resident Inspector Approved by: A C M0+$.e . b 3ll8lSt E. C. McCabe, Chief, Reactor Projects $- . ton IB= Date Summary: Inspectio150-213/88-02(1/13-2/29/87)

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Areas Inspected: This was a routine safety inspection (191 hours0.00221 days <br />0.0531 hours <br />3.158069e-4 weeks <br />7.26755e-5 months <br />) by the resident inspectors. Areas reviewed included outage activities, radiation protection, fire protection, security, maintenance, surveillance testing, events occurring during the inspection period and preparatiens for start-u ,

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Results: No violations were identified and no unresolved items were opene ;

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8803210219 880311 I

PDR ADOCK 05000213

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TABLE OF CONTENTS PAGE Summary of Facility Activities....................................... 1 Plant Operations..................................................... 1 Plant Operations Review Committee.................................... 2 Maintenance and Surveillance Activities.............................. 2 4.1 Reactor Core Support Barrel Thermal Shield Repair........... ... 2 4.2 Spent Fuel Building Sluice Gate Repairs......................... 3 4.3 Cable Vault CO2 System Testing.................................. 4 NRC Information Notice 87-44, Thimble Tube Thinning in Westinghouse Reactors. ..................................... ..................... 4 Events Occurring During the Inspection............................... 5 6.1 Licensee Event Reports and Safeguards Event Reports............. 5 6.2 Steam Generator Blowdown Isolation Valve Modification Errors.... 5 6.3 Containment Penetration Local Leak Rate Testing Failures. . . . . . . . 6 6.4 Inadequacies Identified in Containment Penetration Modification . Periodic and Special Reports......................................... 9 Qualification of Containment High Radiation Monitors................. 9 Eng i nee ri ng Expe rti se On Shi f t. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 10. Security Badge Retrieval........... .. .... .... .................... 10 l

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11. Exit Interview........ .............................................. 10 i

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DETAILS

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1. Summary of Facility Activities t

During this inspection period the licensee completed modifications to the core barrel and thermal shield. Preparations for restart were also initiated in-cluding reperformance of selected surveillance tests. Tests selected included those which would come due before the next outage and those more complex tests which involve interaction of many plant systems. Reactor core reload began on February 20 and was completed on February 2 Mode 5 was entered on February 2 . Plant Operations The inspector observed plant operation during regular tours of the following plant areas:

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Control Room --

Security Building

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Primary Auxiliary Building --

Fence Line (Protected Area)

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Vital Switchgear Room --

Yard Areas

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Diesel Generator Rooms --

Turbine Building

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Control Point --

Intake Structure and Pump Building Control room instruments were observed for correlation between channels and for conformance with Technical Specification requirements. The inspector observed various alarm conditions which had been received and acknowledge Operator awareness and response to these conditions were reviewed. Control room and shift manning were compared to regulatory requirements. Posting and control of radiation and high radiation areas was inspected. Compliance with Radiation Work Permits and use of appropriate personnel monitoring devices were checked. Plant housekeeping controls were observed, including control and storage of flammable material and other potential safety hazards. The inspector also examined the condition of various fire protection system During plant tours, logs and records were reviewed to determine if entries were properly made and communicated equipment status / deficiencies. These records included operating logs, turnover sheets, tagout and jumper logs, process computer printouts, and Plant Information Reports. The inspector observed selected aspects of plant security including access control, physical barriers, and personnel monitorin In addition to normal working hours, the review of plant operations was conducted during the following midnight shifts, weekends, and holidays:

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January 30, 1988, 2:00 PM to 4:00 PM

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February 15, 1988, 6:00 AM to 10:00 AM

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Februa ry 28, 1988, 12:00 Noon to 3:30 PM No unacceptable conditions were identified. Operators were alert and dis-played no signs of inattention to duty or fatigu l

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3. Plant Operations Review Committee (PORC)

The inspector attended several Plant Operations Review Committee (PORC) meet-ings. Technical specification 6.5 requirements for required member .ctendance

were verified. The meeting agenda included procedural changes, preposed changes to the Technical Specifications and field changes to desi'a change packages. The meeting was characterized by frank discussions ar; questioning of the proposed changes. In particular, consideration was giv n to assure clarity and consistency among procedures. Items for which adttuate review time was not available were postponed to allow committee membe's time to re-view and commen Dissenting opinions were encouraged.

I The inspectors have noted that, immediately following the daily norning meet-ing, a short meeting with the department supervisors is routinely held to re-

! view Plant Information Reports (PIRs). PIRs are a means of informing station management of any station conditions, significant equipment failures, or events. This includes matters which may be reportable to NRC. PIRs are generally reviewed by management within one working day of occurrence. At this time reportability to NRC and the need for follow-up or additional ac-tions are evaluated. The inspectors have observed thorough reviews of PIRs and also noted that the PIR process has proven an effective method of bringing management attention to station condition . Maintenance and Surveillance Testing The inspector observed various maintenance and problem investigation activi-ties for compliance with requirements and applicable codes and standards, QA/QC involvement, safety tags, equipment alignment and use of jumpers, per-sonnel qualifications, radiological controls, fire protection, retest, and reportability. Also, the inspector witnessed selected surveillance tests to determine whether properly approved procedures were in use, test instrumenta-tion was properly calibrated and used, technical specifications were satisfied, testing was performed by qualified personnel, procedure details were adequate, and test results satisfied acceptance criteria or were properly dispositione The following activities were reviewed:

4.1 Reactor Core Support Barrel Thermal Shield Repair The repair and modification to the reactor core support barrel thermal shield attachments was completed during this inspection period. A set of six keyways were installed to the upper rim of the thermal shiel This modification was made to limit the thermal shield tangential motion which was responsible for the previous failure of the thermal shield to core support barrel fasteners. These activities have previously been discussed in NRC Inspection Reports 50-213/87-25, 87-27 and 87-31. Ad-ditionally, the licensee has provided information on t'e repair program by letter dated February 25, 1988 (Serial B12808).

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.The inspectors observed the work on the core support barrel assembly on a day to day basis during the inspection period. This included coverage of diving evolutions and milling machine operation. The inspectors found that the work activities conformed with the procedures, extensive Quality Assurance surveillance coverage was conducted and the work received a high level of licensee management oversigh The underwater diving made to perform manual evolutions required by the repair process was of special interest to the inspector Radiological protection of diving personnel was initially examined in NRC Inspection 50-213/87-25 during the initial dives. Inspection coverage of diving evolutions continued through the current inspection period. The inspec-4 tors routinely observed pre-dive surveys of the work area, the placement

] of barriers which prevented the diver from leaving the designated work area, the placement of personnel dosimetry, the pre-dive briefings, l radiological safety and remote dosimetry monitoring during the dives, i and the control of contamination as divers or tools and equipment exited the reactor cavity poo The inspectors found that this specialized aspect of the Radiological Protection Program remained strong throughout the period of the extended

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outage. During each dive observed there was good coordination between those directing the work evolution, the dive support personnel and the supervising Health Physics Technician. In every instance it was obvious that prime responsibility for diver safety rested with the diver support

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personnel and that radiological safety rested with the station personnel.

L There were no unacceptable conditions identified.

l 4.2 Spent Fuel Building Sluice Gate Repairs j During preparations for core reload, the licensee identified that the j spent fuel pool sluice gate valve would not fully close easily due to interference between the valve shaft and shaft guide bushings. The

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. sluice gate valve is on the Fuel Building side of the fuel transfer tube i

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and is used for isolation of the spent fuel pool from the refueling cavity should there be a loss of cavity inventory. The valve is a

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Chappman gate valve with a long shaft which is guided by three bushing assemblies attached to plates in the spent fuel pool liner. This gate valve is air operated to open and close, however, under emergency situ-ations, it is designed to gravity close. During refueling preparations j with the transfer tube gate valve (valve on the containment side of the j transfer tube) closed, the licensee had cycled the sluice gate valve .

several times. It was during these evolutions that the licensee identi-

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fied that the valve would not close unassisted. Underwater videos of valve operation were taken using the mini-submarine utilized for the core j support barrel thermal shield repair From these videos, the licensee

determined that the valve shaft was slightly bent and its movement was

restricted by the lower bushing assembly. Through the Jumper / Lifted l Lead / Bypass process, the licensee analyzed and PORC approved the disen-l gaging of the lower bushing assembly. This assembly was then lowered

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to rest on the valve gate. With this bushing bypassed, it is possible for the shaft to bend further if the valve were to be closed by ai To prevent further shaft degradation, the licensee hung, on the valve -

controls, a sign stating that the valve is to be closed solely by gravity.

1 The valve was successfully stroke tested several times; air to open and gravity to close. Inspector review of these activities and the bushing disengagement identified no deficiencie .3 Cable Vault CO2 System Testing j

'l On February 23, the inspectors observed licensee conduct of SPL 10.7-336, Containment Cable Vault CO2 Suppression System Concentration Test. The '

purpose of this test was to demonstrate that the cable vault CO2 system conforms to the original design requirement of providing a CO2 concen- *

tration of 30% within two minutes and 50% within seven minutes on both levels in the vault, This test was conducted because the licensee was I unable .to obtain original station records detailing previous test results.

, During test performance, extensive personnel protective actions were i

taken due to the danger involved with this test. The test was performed successfully, however evaluation of test data indicated that the CO2- .

concentrations in the lower level fell slightly below the acceptance criteri Because of this, the CO2 system was declared inoperable on

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February 27. In accordance with Technical Specification (TS) 3.22.B.2, a continuous fire watch was established. The TS also requires that a backup fire suppression system be provided; the licensee has provided a local hose station in the Service Building hallway and can manually discharge both the normal and backup CO2 banks should there be a fir Currently, the licensee is evaluating methods to ensure a sufficient

, concentration of CO2 is provided. These corrective actions will be re-l viewed during future inspection ;

5. NRC Information Notice 87-44, Thimble Tube Thinning in Westinghouse Reactors

- This Information Notice (IN) was issued to alert licensees to potential prob-

! lems with thimble tube thinning in the area between the lower core plate and '

j the bottom of the fuel assembly guide tubes. The licensee's actions in re-sponse to this notice were previously discussed in NRC Inspection Report 50-213/87-3 ,

) In response to this notice the licensee performed eddy current testing (ECT)

on the thimble tubes. Three tubes were identified as having wear in the area

of concer Two tubes displayed wear of 20% and one had wear of 30%. The .

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wear has been characterized as localized pitting which is different from the t i fretting observed at other nuclear plant The licensee has elected not to i

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take any immediate actions based on these observations. However, ECT will i

be performed on the tubes in the next refueling outage. At that time, the new data will be evaluated and compared to the observations from this outag Any necessary corrective actions will be determined at that time.

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6. Events Occurring During the Inspection Licensee Event Reports [LERs and Safeguards Event Reports (SERs)]

The following LERs and SERs were reviewed for clarity, accuracy of the descript!on of cause, and adequacy of corrective actic... The inspector determined whether further information was required and whether there were generic implications. The inspector also verified that the report-ing requirements of 10 CFR 50.73, 10 CFR 73.71, and Station Administra-

, tive and Operating, 4ad Security Procedures had been met, that appro-priate corrective action had been taken, and that the continued operation of the facility was conducted within Technical Specification Limits.

I 88-01 Design Error Found In Steam Generator Blowdown Isolation Circuit l-

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88 02 Fire Detection Subsystem Declared Inoperable Due to Damaged Heat Detectors No unacceptable conditions were identifie .2 Steam Generator Blowdown Isolation Valve Modification Errors j On January 14, during licensee evaluation of station drawings as part i of the update / verification of the Master Equipment and Parts List, it

was identified that an error existed in a design change made [ Plant De- l sign Change Record (PDCR) 362] to the Containment Isolation System in 1980 as part of the post-TMI modifications. This change was in response to NRC concerns about equipment automatic actions after an Engineering i Safety Features System reset. Specifically, the Steam Generator Blowdown isolation valves (BD-TV-1312-1, 2, 3, and 4) would not trip closed in I response to a High Containment Pressure (HCP) signal preceded by a loss i

of voltage on 4160V buses 1-2 and 1-3 or 480V buses 4, 6, or 7. Addi-

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tionally, these four valves would not have closed upon receipt of an i undervoltage signal on either of these buses. LER 88-01 was submitted and PDCR-930, Steam Generator Blowdown Isolation Circuitry Changes, was .

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created to correct this condicion. The PDCR relocated the logic for the

- bus undervoltages to be in series with the HCP trip logic and modified I the relays to correspond to th? de-energize to open actuation logic.

) Post-modification resting was successfully performed under SPL 10.7-335, Preoperational Test of PDCR-930, to verify that the circuitry is now k

correcte The licensee revieweJ other containment isolation valve trip circuitry mcdified by PDCR-362 to verify that the other valves would function as

designed. No further deficiencies were found. Also a task force was formed to review post-TMI modifications involving electrical design

changes to verify correctness of the design and its implementation. The

task force reviewed the design requirements, the functional impact of

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the design change on the modified system (s), and the design drawings to verify that the requirements were met. No further deficiencies were identifie The inspector had no further questions on this matte .3 Containment Penetration Local Leak Rate Testing Failures Due to the long duration of the current refueling outage, the licensee elected to reperform several surveillance tests to minimize the potential need for a mid-cycle shutdown for overdue surveillances. The repeated surveillances included snubber testing, system functional testing, and containment isolation valve local leak rate (LLRT) testin LLRTs were not required to be redone because the plant had not left Mode 5 since they were last performed. However, since containment integrity has been a sensitive issue, these tests were included. Eleven contain-ment penetrations which have historically exhibited high as-found leak rates and repetitive LLRT failures were chosen. Four of these valves are for the reactor coolant pump seal suppl LLRTs on these penetra-tions would result in unnecessary personnel exposure. For this reason these penetrations were not retested. Of the seven penetrations tested, two failed: P-65, Containment Air Sample; and P-71, Primary Vent Heade In addition, P-80, Auxiliary Spray From Fire System, was observed to be leaking after successful stroke testing; it subsequently failed a LLR P-71, Primary Vent Header, failed the penetration LLRT with leakage of 429 lb-mass / day, exceeding the Inservice Inspection (ISI) acceptance criteria of 5 lb-mass / day at 40 psi Technical specification (TS) requires total containment leakage to be less than 650 lb-mass / day. At the beginning of the refueling outage (July l'?") this penetration was tested and exhibited a leakage of 0.16 lb-ma. Jay. During station out-ages only, this penetration is used to vent a;r from the primary plant during fill and vent evolution The penetration has two in series car-bon steel, angle globe stop valves which serve as containment isolatien valves, VH-V-522 inside containment and VH-V-525 outside containmen Both of these valves are kept locked closed during station operatio During performance of the LLRT, operations personnel noted a small water discharge when the test connection valve (VH-V-523) was opened. The LLRT ( was performed, in accordance with SUR 5.7-74, Primary Vent Header (P-71),

i with unsatisfactory results. To determine the leakage path, the test equipment was relocated to an alternate test connection between the isolation valves (VH-V-524). Leakage was determined to be 170 lb-mass /

, day. This reduced leakage suggests that test boundary valve VH-V-521 l greatly contributed to the initial leakage measurement. Both penetration l isolation valves were cleaned and the system was thoroughly flushed with

! primary water. A second LLRT was performed with unsatisfactory results, l 798 lb-mass / day. The majority of this leakage was determined to be through the outboard boundary valve (VH-V-521). This valve was also disassembled and cleaned. A satisfactory as-left LLRT was then performed l

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with leakage measured at 0.05 lb-mass / day. The licensee concluded that normal operation of the vent system during the outage resulted in the

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passage of Reactor Coolant System fluid through these valves. Any sedi-ment present could easily have become lodged in the valve seat / disc area

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because these valves are installed upside down (VH-V-525) and sideways (VH-V-522) in the lines. To ensure minimal penetration leakage when containment integrity is required, the licensee has established a per-manent start-up LLRT to be performed after completion of fill and sent evolutions and before containment integrity is set. The valves would then be locked closed for the duration of the operating cycle.

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The Containment Air Sample penetration (P-65) also failed its LLRT retest with leakage 4.i 19 lb-mass / day (its ISI acceptance criterion is 10 lb-

mass / day). Tbc as-left LLRT performed in July 1987 was 0.17 lb-mass / da Two check valves, VS-CV-1103 and 1104, one on cach side of the contain-ment wall, serve as containment isolation valves for this penetratio The as-found test performed in July 1987 identified 41 lb-mass / day leak-age from VS-CV-1104 and no leakage from VS-CV-1103. VS-CV-1104 was dis-assembled and the presence of loose dirt and dust was identified. The valve was not reworked; it was reassembled and the LLRT retest was satis-

[ factor This most recent LLRT identified a leakage of 19 lb-mass / day

! for VS-CV-1103 and no leakage for VS-CV-1104. When VS-CV-1103 was dis-l assembled and inspected, no seat or disc degradation or dirt accumulation was observed. The valve was reassembled and retested satisfactorily with a leakage of 0.61 lb-mass / day. The licensee concludec that dirt or dust may have been lodged between the valve seat and disc but fell away when j the valve was disassembled. Currently, the licensee is evaluating

, several possible long-term corrective actions to prevent future LLRT failures for this penetration. These potential solutions include in-

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stallation of heat tracing to eliminate buildup of condensation and altering the check valve design.

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P-80, Auxiliary Spray From Fire System, failed its LLRT under different circumstances. This penetration provides river water from the Fire Pro-l tection System to the containment spray header and would only be used under extreme emergency conditionc. After successfully stroke testing MOV-RH-31 in a routine operability verification, operations personnel noted unexpected water leakage from the penetration crain valve, RH-V-3 Under Engineering recommendation, a LLRT was performed. The penetration exhibited a leak rate of 1237 lb-mass / day. This alone exceeds the TS 4.4 limit for total containment leakage of 650 lb-mass / day. The valve was disassembled and inspected. An accumulation of silt and sand was identified between the valve seat and disc. The valve was cleaned, re-assembled and successfully leak tested (1.5 lb-mass / day).

This valve has been leak tested three times during this refueling outag l The In-Service Testing (IST) Program requires that it be stroke tested during each cold shutdown and, when outages are extended as this one was, quarterl The penetration was successfully leak tested in July 198 The first quarterly stroke test was successfully performed in November

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1987; leakage was observed from the drain line. When the second quar-terly stroke test was performed in February 1988, similar leakage was again observed. In both cases, after flushing and cleaning, the pene-tration passed the LLRTs. This valve is not onerated for purposes other than operability verifications because this would result in a spray down of containment. Containment integrity can be assured by LLRT after valve manipulacion. The licensee is currently evaluating alternative testing sequences to preclude LLRT failures after operability stroke testing.

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Licensee maintenance and testing activities in response to these LLRT failures was thorough. Long-term corrective actions for P-80 and P-65 will be reviewed during future inspection .4 Inadequacies Identified In Containment Penetration Modifications During this refueling outage, the licensee made changes to several con-tainment penetrations under Plant Design Change Record (PDCR) 878, Ap-pendix A and J Penetration Modifications. Included in this modification was the installation of nine solenoid-operated valves (50Vs). The ori-ginal PDCR specified SOVs with Elastomer seating surfaces. After the purchase order for these valves was processed, the licensee determined it prudent to substitute SOVs with Viton seat (The Elastomer seats tended to heat up when energized for long periods and then had the potential to stick in the closed position.) The vendor supplied kits for the seat replacements after the purchased valves were installe During this valve seat changing process the licensee identified defi-ciencies with the initial valve installatio The PDCR had been performed by a contractor (C.N. Flagg) earlier in the outage. When site maintenance personnel were changing the valve seats, they identified an error in the size of terminal lugs which were in-stalled. Furthermore, wire braiding was not removed before installation of the Raychem splices. This was brought to management attention through the Plant Information Report (PIR) process. All nine valves were removed for seat replacement and the deficiencies were correcte Generation Construction has informed the contractor of these deficiencie Preliminary information indicates that contractor supervisors were aware of the correct methods for valve installation but that work crews did not adequately follow procedures. Contractor Quality Control (QC) per-sonnel were also used for this PDCR. The licensee has indicated that the contractor is performing a review of their QC program and coverage and personnel procedure adherent.e. An inspection of work done by this contre.ctor has been performed and no further deficiencies were identifie The inspector had no additional concerns at this time; this matter will be reviewed in future inspection <

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. Periodic and Special Reports

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Upon receipt, periodic and special reports submitted pursuant to Technical Specification 6.9 were reviewed. This review assessed whether: the reported information was valid and included the NRC required data; test results and supporting information were consistent with design predictions and performance specifications; and planned corrective actions were adequate for resolution of the problem. The inspector also ascertained whether any reported informa-tion should be classified as an abnormal occurrence. The following p;rindic reports were reviewed:

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Monthly Operating Report 87-12, Covering the Period December 1, 1987 through December 31, 198 Monthly Operating Report 88-01, Covering the Period January 1 through January 31 198 New Switchgear Building Construction Bimonthly Progress Report No. 8, dated February 1, 198 . Qualification of Containment High Radiation Monitors In February 1987, the NRC and several licensees (including Northeast Utilities)

received a 10 CFR Part 21 report from General Atomic Technologies, Inc. in-forming them that a defect was identified in the Sorrento Electronics Con-tainment High Range Radiation Detectors. The defect is in the inside con-tainment portion of the ion chamber's signal coaxial cable which was manufac-tured by Rockbestos Co. Under post-LOCA conditions, with elevated containment temperatures, the cable insulation resi tances var The variances are a function of the cable configuration and the containment post-LOCA temperature profil Haddam Neck has two of these detectors (CD-1 and 2) located on the charging floor in containment. They are used for radiation monitoring of containment under accident conditions. Technical Specification (TS) 3.23, Post Accident Instrumentation, requires that two containment high radiation monitors be operable during Modes 1-4, with the capability to measure up to 1E8 R/hr and an alarm setpoint less than or equal to 100 R/hr. Indications from these detectors are used in the Emergency Operating Procedures (EOPs) and Emergency Implementing Procedures (EPIPs). The E0Ps contain a decision making step based on the indications of CD-1 and 2 in ES 1.4, Transfer To Two Path Recir-culation. Whether the detectors are indicating above or below 20,000 R/hr determines if the recirculation path includes systems outside containmen At radiation levels above 20,000 R/hr, the path will not include equipment outside containment. CD-1 and 2 indications are used in EPIP 1.5-7, Radio-

logical Dose Assessment, for offsite dose calculations. The action level for these calculations is when the detectors are reading at least 34,000 R/hr.

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Between February and June 1987, the licensee had extensive communications with Sorrento Electronics to determine the effect of this defect on Haddam Neck's detectors. An analysis was conducted specific to the station's cable con-figuration and post-LOCA temperature profile. These calculations were finai-ized during this inspection period and are documented for the Millstone Units and Haddam Neck in Reportability Evaluation Form 87-2 The licensee con-cluded that CD-1 and 2 exhibit an error of 9.4 R/hr in the nonconservative direction. The licensee has evaluated the implications of this inaccuracy and determined that, since the detectors' ranges are so broad and negligible in the action range, they are operabl It was also noted that the error in the low range does not meet the guidance in Regulatory Guide RG-1.97, instrumentation for Light-Water-Cooled Nuclear Plants to Assess Plant Conditions During and Following an Accident. This is because they do not meet the factor of two accuracy specified for the lower range. The licensee plans to submit an information letter to NRC describing this exception to RG-1.97. This matter will be reviewed during future in-spection . Engineering Expertise On Shift On January 21, the Office of Nuclear Reactor Regulation (NRR) informed the inspectors of concerns involving the use of dual-role Shift Supervisor / Shift Technical Advisors (SS/STA) at the Millstone plants and Haddam Neck. The current Haddam Neck Technical Specifications (TS) Table 6.2-1, Minimum Shift Crew Composition, specifies a crew of seven people during modes 1, 2, and Because the SS serves in a dual capacity (SS and STA), these sevan positions have been filled by six individual This issue is discussed in detail in the Villstone Unit 1 NRC Inspection Report 50-245/88-0 No NRC field in-spection action is planned until the latest licensee's submittal on this matter is evaluated by the NRC.

l 1 Security Badge Retrieval The licensee has installed equipment to prevent unauthorized removal of security keycard Each keycard is tagged with a microwave resonance device similar to those used to prevent theft from retail establishment The de-tection equipment is located at an exit poin This equipment is not a re-quired plant physical protection system but should improve the overall imple-mentation of the program by reducing the likelihood of unauthorized removal of keycards and also assist with the monitoring by security officer 'l . Exit Interview During this inspection, meetings were held with plant management to discuss

the findings. No proprietary information related to this inspection was I

identified.

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