IR 05000255/1995009

Insp Rept 50-255/95-09 on 950704-0821.No Violations Noted. Major Areas Inspected:Operations,Engineering,Maint & Plant Support

ML18065A130
Person / Time
Site:	Palisades
Issue date:	09/15/1995
From:	Kropp W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML18065A129	List: IR 05000255/1995009 ML18065A128
References
50-255-95-09, 50-255-95-9, NUDOCS 9509270197
Download: ML18065A130 (15)

Text

U.S. NUCLEAR REGULATORY COMMISSION REG ION II I

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REPORT N /95009 FACILITY Palisades Nuclear Generating Plant

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LICENSEE-Pa l i sades Nuclear Generating Plan *27780 Blue Star.Memorial Highway Covert, MI 49043-9530 DATES July 4 through August 21, 199 INSPECTORS D. Passehl, Resident Inspector F. Maura, DRS Inspector J. Schapke~. DRS Inspector J. Lennartz, DRS Inspector R. Lerch, DRS Inspector R. Bywater, Resident Inspector APPROVED BY Date AREAS INSPECTED A routine, unannounced inspection of operations, engineering, maintenan~e~ and plant support was performed; Safety assessment and quality verification attivities were routinely evaluate PDR ADOCK 05000255 Q

PDR

• OVERALL ASSESSMENT OF PERFORMANCE Performance during the refueling outage overall has been satisfactor Early in the outage there.was a series of personnel errors that collectively indicated a weakness in the process for controlling plant activities in the areas of communications and attention to detail on the part of plant personne Manage~ent recognized these errors as precursors and took action to improve performanc Improvement began to show later in the outag The licensee generally maintained an improved posture this inspection perio Management oversight of the steam generator inspection program, fuel handling, containment closeout, and plant startup activities was positiv No problems were noted during reactor ves~el reassembly activities, and no mobile crane errors.were note (Problems in these areas were discussed in the previous inspection report 255/94008.)

However, a significant weakness that remained from the previous inspection was in the area of control of foreign materia Management acknowledged this weakness and has begun action to improve performance in this are Planning and coordination of major projects was goo However, the challenge fcir management was to better plan and coordinate the lesser project Opportunities were missed to correct some minor longstanding maintenance deficiencies on the emerg~ncy diesel generator~ following major outages on both machine Two inadvertent safety injection system actuations occurred during testing because of weaknesses in planning and coordination and poor worker practic Problems with plant labelling contributed to two other events during this inspectio False. labelling on instrument root valves led to a reactor trip in Hot Standb Lack of adequate labelling on a hose led to a health physics worker spraying a small amount of primary coolant pump oil into the reactor cavity..

Management's initial response to an unexpected control rod motion event, and to the discovery of the loss of the containment high pressure trip function was satisfactor *

ASSESSMENT BY FUNCTIONAL AREA Performance within the area of OPERATIONS was good (section 1.0).

Uperatio~s' control of control room activities was goo Control room operators were not overburdened with outage related activities and traffic in the control room was maintained at an appropriate level. Control room manning exceeded minimum requirements. Operator distractions were minimized by maintaining the status board and communications center for core offload and reload in the shift supervisor's office, and by including a training department member to augment the day shift cre Communications during refueling activities were appropriate.

Management and engineering's involvement during fuel handling activities was goo Use of procedures and log keeping was adequat However, refueling equipment problems caused unanticipated delays with core reloa Containment closeout inspection was favorabl Control of debris was very good with virtually.no loose material present. This was a large improvement from the outag Only approved items were left in containment and these were appropriately secure Personnel performing the inspection were thorough and followed a procedure that provided adequate guidanc Containment sump debris screens were clean and in good conditio The remainder of the sump was also free of debris except for some sludge buildup on the floor, which the licensee periodically clean Operators' response to a swagelock fitting leak on primary coolant system flow transmitter DPl-011288 was goo However, labelling weaknesses were identified on a few primary system instrument root isolation valve The. performance of operators was good during the appro~ch to critical operation Good communications and professipnalism were exhibite An unexpected control rod motion event occurre Operator re~ponse was appropriate. This issue will be assessed in more detail in the next routine inspection repor Performance within the area of MAINTENANCE was adequate (section 2.0)..

Maintenance perform~d many activities well and without significant error Examples included numerous pump and valve repairs, seal replacements on three primary coolant pumps, switchgear preventive maintenance, and numerous emergent work issue However, weak portions of some maintenance procedures, and some poor worker practices were noted. Opportunities were missed to _

correct some minor maintenance deficiencies on the emergency diesel generators during maintenance outage *

Weaknesses were noted in control of foreign material.. Two examples were note One was when a health physics worker inadvertently sprayed less than a gallon of oil into the reactor cavity while attempting to rinse the upper

guide structure. This hose had earlier been used to make adjustments to the primary coolant pump oil syste The hose had not been drained of its.

residual oil and was not labeled. Another example was when a plastic bag for waste radiological materials was found in safety injection pump P-66 Management acknowledged' that plant procedures governing FME could be improved and that actions have been initiated to revise appropriate procedure Management expectations we~e not met during perfor~ance of a safety injection system test for containment high pressure and containment spra Two separate inadvertent safety injection actuations occurred during the tes The first was due to a weakness in the surveillance test procedur The second was due to personnel erro One contributing reason was that the test' steps where the errors occurred were new portions of the test that had not received an adequate review by persons performing the tes ;,

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A major overhaul of emergency diesel generator (EOG) 1-1 was conducted as a result of enforcement action taken for failure to perform vendor-recommende maintenance in accordance with Technical Specifications (see inspe~tion report 255/94017).

Most engine components appeared in satisfactory conditio Two significant observation~ were evidence of engine exhaust recirculating into the air intake of the turbocharger, and an indication of water leakage into cylinder 7 Both of these* items were considered inspector followup item Some lesser maintenance was performed on EOG 1-2.as wel Several minor work requests were still outstanding at the conclusion of the work on both EDG Planning and coordination between Maintenance and other departments to address these minor items warranted improvemen Performance within the area of ENGINEERING was adequate (section 3.0).

The licensee discovered that the reactor protection system logic trip function for containment high pressure was inoperable during the last two operating cycle Among the apparent root causes was an inad~quate post modificatio test during an upgrade of this portion of the reactor protection system back in 1992.. A detailed review of this event was performed and documented in special NRC inspection report 255/95010(DRS).

The steam generator (SG) eddy current *inspection program was conservativ The licensee inspected 20 percent of the SG's tubes versus a 3 percent requiremen Additional testing using the motorized pancake coil and plus point probe was also performe These examinations were performed at the to of the tubesheet area where recent industry findings identified *

~ircumferential cracking (Reference Generic Letter 9~~03).

No SG tubes required repai Planning efforts for performance of reactor vessel annealing were thorough and in compliance with NRC Draft Regulatory Guide 102 Examples of good planning efforts include use of mockups to demonstrate feasibility and reduce exposure-for placement of thermocouples within the annulus; and selection and placement of pressure vessel materials in surveillance capsules located in high neutron flux regions to get add it i ona 1 information on the amount of recovery for

annealed vessel materials:

As part of a Regional Request, applicability of Information Notice 94-66, Supplement 1, 110verspeed of Turbine-Driven Pumps Caused by Binding in Stems of Governor Valves, 11 was reviewed for Palisade The licensee has not had any experience with governor valve stem binding or corrosion problems associated with its* turbine-driven auxiliary pum Pump components are inspected during routine preventive maintenance activities and pump performance is monitored

• during routine surveil 1 ance testin *

Performance within the area of PLANT SUPPORT was adequate (section 4.0).

The main steam line penetration through the turbine building/auxiliary building wall contained an opening that was greater than the allowable size for an opening in a vital area barrier. A security officer was posted at the opening as an immediate compensatory measure and a condition report was initiated to document the finding and corrective actio The safety

significance of the opening was minor because of its ob~cure location, small size, and environmental conditions due to proximity to the hot steam ltn Management later determined that the opening was not a viable pathway and relieved.the security office Summary Of Open Items Ins~ector Follow-up Items: identified i~ section Unresolved Item: identified in section 2.1.

• INSPECTION DETAILS OPERATION NRC Inspection Procedures 71707 and 92709 were used in the performance of an irispection of ongoing plant operation The findings showed performance was goo * Control Room Observations Control room operators were not overburdened with outage related activities, and traffic in the control room was maintained at an appropriate leve Control room manning during the outage consisted of the same.number of licensed operators that would be on shift during power operations. This exceeded shift manning requirements during refueling and was considered a strengt Palisades implemented a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> rotating shift schedule during the outag Each crew's first shift following off days was spent off watch reviewing plant conditions, status of outage work, and preparing for future work activities. This was considered a strengt To reduce distractions in the control room, the status board and communications center for refueling operations were maintained in the shift supervisor's offic The inspector concluded that communications regarding -refueling activities were appropriat A training department member was aisigned to augment the day shift cre This reduced distractions on th~ crew and was considered a strengt.2 Fuel Handling Observations Core reload was delayed several days due ta problems with the refueling

~achine.. The licensee made the appropriate repairs and successfully tested the refueling machine before core reload was commence Fuel handling activities were further delayed due to failure of the fuel handling transfer cart winc Other related observatio~s were:

Proper communications were established between personnel in containment and the fuel handling buildin Operations management oversight was observed in containment and *

the fuel handling building.

Reactor engineering and system engineering oversight was present during fuel move *

Appropriate procedures were used and logs maintained during the fuel move.3 Containment Closeout Inspection Containment cleanliness deteriorated during the outag There were trip hazards on the refueling floor and a lack of fall protection on the west side of the refueling cavity inside of the debris free zon To correct this, supervisors were assigned to tour containment and look for problem Reminder notes were placed in the daily bulletin~ As the outage concluded, extraneous material was removed from containment and conditions improve *

The inspector's assessment of a containment closeout-inspection was favorabl Control of debris was very good with virtually "no loose material presen Only approved items were left in containment and

• these were appropriately secure Personnel performing the inspection

-~, were thorough and followed a procedure that provided adequate guidanc 'Two minor items noted by the inspector were peeling safety tape on some ladd~r rungs and a foam bench seat mounted on the fuel manipulator crane platfor The inspector's observation of a post refueling videotape of the containment sump indicated that the debris screens were clean and in good conditio The sump itself was free of debris except for some sludge buildup on the floo Plant personnel stated that this sludge is removed at periodic interval.4 * Manual Reactor Trip During Control Rod Drop* Timing Test Du~ To A Leak On A Primary Coolant System Flow Instrument On August 15, 1995, operators manually tripped the reactor when an 11 gpm leak started at*a swagelock fitting on primary coolant syst.em flow transmitter DPl-0112B At the time of the leak, the plant ~as in ~ot Standby and all control rods (with one exception) were latch~d and at

• the 2.5 inch po~ition. One control rod was being withdrawn to perform rod drop time testin An operator was dispatched to close the rotit valves for DPI-0112BB to stop the lea However, the operator closed the incorrect valve due to a lab~lling error. This resulted in a "A" channel RPS low flow pre-trip alar Based on the indications of multiple RPS channels tripping, the shift supervisor ordered a manual reactor scra The cause of the leak was due to incorrect installation of the swagelock fitting several years ag The licensee checked a large sample of other similar swagelock fittings and found no other installation problem In addition, no labelling discrepancies were found on local isolation valves for other instrument However, the licensee found other labelling discrepancies on a few other root isolation valves located on the primary system pipin Because the root

• valves represent personnel hazards from a dose and temperature perspective, and were inaccessible at power, the licensee would correct the labelling during the next cold shutdown opportunit.5 Initial Cycle 12 Criticality On August 17, 1995, pl~nt operators made the reactor critical for the fiist time in cycle 1 The inspector observed that the estimated critical boron concentration was within the predicted target ban The performance of the operating crew was goo There were good communications and professionalism exhibited during the evolutio The plant again went critical on August 19, 1995, following a brief shutdown to make repairs to the rod control system (see bel-0w). Control Rod Withdrawal When Given An Insertion Demand On August 17, 1995, during low power physics testing following initial criticality for cycle 12, an unexpected control rod motion occurre Group 4 rods had been pulled as part of the first phase of low power physics testin When reinserting the group 4 rods to their original position, grriup 4 rods no. 40 and no. 41 both moved out upbn an insertion deman Rod control was in manual sequentia Plant operators placed rod control to manual individual and rod no. 41 inserted; however, rod no. 40 continued to rise when given an insertion deman The 'licensee declared rod po. 40 inoperable and commence boration to subcritical conditions (critical boron concentration with all rods out plus 100 ppm).

Plant operators later tripped the reactor when initial troubleshooting of the as-found condition identified a short circuit between the lower and raise control rod circuitry. This issue will be assessed in more detail in the next routine inspection repor.7 Other Observations During plant heatup the licensee exceeded the administrative requirement for maintaining containment building air pressure below one psi The

• ~aximum pressure attained was just over 1 psi The technical specification limit is less than 3 psi The licensee was unable to use the normal vent path due to a blockage in the release path. This blockage is located in* the vent gas collection header (VGCH) and is one of the oldest and most problematic operator workarounds being tracked by

• the 1 icense As a result, the 1 icensee implemented a temporary modification to reroute the containment vent path around the blockag The licensee has a plan to locate and attempt to free the blockages during the next few month.0 MAINTENANCE NRC Inspection Procedures 62703 and 61726 were used to perform an inspection of maintenance and testing activities. The findings showed maintenance was adequat '

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• weaknesses In Foreign Material Control As indicated in the last inspection report, there w'ere weaknesses noted in the control of foreign material within the debris free zones of the reactor cavity and spent fuel poo Two additional examples of weak FME control are detailed belo The two procedures governing foreign material exclusion were MSM-M-20, "Maintenance Cleanliness Guidelines" and MSM-M-47, "Foreign Material Exclusion in the Spent Fuel' Pool Area and Reactor Cavity".

The lice~see acknowledged that these plant procedures governing FME could be improved and that actions have been assigned to improve in this ar~a. Pending completion of the licensee's actions, the issue of foreign material control is considered an unresolved item (255/95009-01 (DRP)).

2.1.1 Oil Inadvertently Sprayed Into Reactor Cavity A health physics ~orker inadvertently $prayed less than one gallon of oil into the reactor cavity while attempting to rinse the upper guide

- structur ln preparing to spray the upper guide structure with flushing water, workers connected a hose that had been stowed in containment to the water source used for washdow However, this hose had earlier been used to remove oil from the reactor coolant pump The hose had not been drained of its residual oil and was not labele When the water supply was turned on, it flushed the residual oil into the reactor* cavit *

Chemistry personnel promptly sampled the cavitj wate The results showed no deleterious effects. While as little as 250 ml of oil would disrupt the water chemistry, samples routinely taken of primary coolant water have not shown any evidence of oil. The identification and response to the problem by Operations and Chemistry was timely and thoroug.1.2 Plastic B~q Found Inside High Press~re Safety Injection Pump*

Safety injection pump P-66A failed a routine surveillance test as differential pressure was in the required action range and pump vibration was in the alert rang When the pump was disassembled for inspection, a yellow plastic bag was found wrapped around the pump shaft and lodged in the first stage impelle The pump was not required to be operable sin~e reactor was shutdown at the tim Plant workers qegradatio reassemble statu inspected the pump and found no further signs of The rotating element was overhauled and the pump was The pump was satisfactory testing and returned to operable *

The licensee evaluated possible locations where the bag may have entered the system and evaluated operability of the other safeguards pump The inspector reviewed the licensee's investigation and operability of the other pumps and had no further concerns.

• • Inadvertent Safety Injection Actuation An inadvertent safety injecti~n {SI) actuation occurred during performance of surveillance procedure R0-12, "Containment High Pressure

{CHP) and Spray System Tests", Revision 2 A second inad~ertent SI actuation occurred during restoration from this tes One contributing reason was that the test steps where the errors occurred were new.

portions of the test that had not received an adequate review by persons performing the tes The test setup {step 5.4.3) required two wires to be lifted {from relay TVX-L terminal two) to prevent the CHP signal from energizing the safety injection relay The technician lifted these two wires from the terminal as directed, but then taped them together. This allowed the CHP SI signal to be transmitted when the test pressure signal was applie The procedure did not specify separating the two wires onc they were lift~d from the termina All safeguards equipment operated as expected and the surveillance test*

was suspende The second inadvertent SI actuation occurred during recovery from the suspended test. While landing a wire, the technician shorted the circuit when his screwdriver touched the terminal bloc All safeguards equipment again operated as expecte The crew used appropriate procedures to verify proper operation of SI equipment and to subsequently restore the equipment to the required status following both inadvertent SI actuation.3 Emergency Diesel Generator 1-1 Overhaul

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The inspectors witnessed selected portions of the first major disassembly of diesel 1-1 after over 20 years of standby service, which included 3830 hours0.0443 days <br />1.064 hours <br />0.00633 weeks <br />0.00146 months <br /> of operatio Internals of the engine and turbocharger were inspected for evidence of abnormal wea I~ general, the conditinn of the engine and its components was goo The following observations were noted:

The lower half of the connecting rod bearings showed some areas of overlay removal, apparently the result of the hydraulic effect of compressed oil. This was more pronounced on bearings 2R, 3L and 4L. In addition, several round "stains" were noted on 3L. The latter was not consistent with the observations of the other bearings. Non~ of these indications were considered detrimenta Well healed, scuffing marks were noted on the bore of liners IL, IR, 2R, and 3 Similar marks were present in their respective piston. These marks were probably the result of the 1985 overheating event.

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Of more significance was the evidence of engine exhaust recirculation into the air intake. * The turbocharger fan blades were coated with a layer of fine carbon residu This foreign material could reduce the efficiency of the turbocharger and has the potential for creating an imbalanced conditio In addition, the inlet valves of cylinder heads IL and 7R had a thick (1/32" to 1/16") coating of carbon grit. The presence of this foreign material could result in valve sticki-ng and subsequent mechanical failure. The licensee will be considering changes in exhaust or inlet location to prevent these abnormal conditions from recurrin In addition, cylinder head 7R had an indication of water leakage (rust mark on seat).

The licensee will attempt to determine the source of the wate The inspector wi.11 continue to follow the licensee's actions in this are Pending the licensee's evaluation of the engine exhaust recirculation issue, this will be an inspector followup item (255/95009-02 (ORS)).

Also, pending the licensee's evaluation of the water in cylinder 7R, this will be an inspector followup item (255/95009-03 (ORS)).

2.3.1 Emergency Diesel Generator 1-1 Integrated Testing During the July 11 performance of Procedure T-302, "Overspeed Trip Test", the licensee identified that an incorrect overspeed trip spring

• had been installed during the EOG overhaul *as a result of a procurement

• error. The proper replacement was ordered and installed on July 1 The test was then completed satisfactoril.3.2 Other Emergency Diesel Generator Observations The licensee completed maintenance outages on both EOG 1-1 and EOG 1-2 during the current refueling outag Following the maintenance outages,

• the inspector observed several outstanding work requests that had not been addresse Some ~ppeared minor and were over one year ol One example was a work request initiated on April 14, 1994, to replace the belly tank level switch on ED~ 1-2 with a functional equivalen The inspector followed up and noted the licensee had scheduled those ~arious work requests for later this yea Although the licensee performed extensive maintenance on both EDGs during this refueling outage, the licensee acknowledged that opportunities were missed to address some of the long-standing minor issues on the EOG.4 Other Maintenance Observations

WI-TGS-1-008 (Revision 1), Testing of OEH Runback Circuit The responsible engineer provided a very good prejob brief to all personnel involved in the tes Good coordination and control of this activity was also observed.

11 ENGINEERING (37551, 37700, 73753)

NRC Inspection Procedures 37551 and 73753 were used to perform an inspection of engineering activities. The findings showed performance was adequate, however, an example of unacceptable performance was identified in paragraph 3.1 belo * Reactor Protection System (RPS) Logic Trip Function For Containment High Pressure Was Inoperable On July 28, 1995, while performing RPS circuit checkouts after connector replacements, the licensee discovered that the RPS. logic trip function for containment high pressure was inoperable. A detailed review of this event was performed and documented in special NRC inspection report 255/95010(DRS).

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. Steam Generator (SG) Inspection Performed Appropriately The NRC inspector observed portions of SG eddy current testing. Testing was performed in accordance with commitments to Generic Letter 95-0 Testing was conservative and exceeded requirements~

Review of the licensee's inspection program, procedures, analysis guidelines, examiners certifications, and graphics and inspection data concluded that Palisades SG performance is good and the inspection program is conservatiV~. The program utilized state of the art equipment and inspection technique Results of the last three SG inspections have not identified any significant tubing degradatio The majority of the degradation found

.. to ~ate appears to be wear, caused by the tubes' contact with antivibration bars or bat wing support This type of wear is common among recirculating type SGs and reduces with time in operatio No SG tubes required repai.3 Palisades Reactor Vessel Annealing Meeting Summary A meeting was held at Palisades on June 28 to discuss the reactor vessel annealing planned for the 1998 refueling outag The vessel will be annealed using an indirect gas heating syste Ducktwork and exhaust ventilation will be routed through the equipment and/or escape hatc Inside the reactor vessel there will be five heating zones with two independent burners per zon Palisades currently plans on a controlled

• thermal profile during heatup with a 7-day soak at 850°F and a 25°F/h cooldow Temperature inside the rig at the source will be approximately 1400° Consumers predicts approximately an 80 percent or greater recovery from the annealing; however, it estimates only 40 percent recovery is required to reach its target of 2011, which includes recapturing the construction perio NRC staff raised con*cerns over adequate fire protection planning for onsite storage of the large quantities of

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r t *combustible gas required for annealin Walkthroughs of systems, structures and components within containment, potentially affected by the proposed annealing, were conducted with NRC staf The content and schedule for the Application for Approval for Thermal Annealing (AFATA) was di~cusse NRC staff raised concerns over*

the. content and timetable proposed by Consumers Power Company (CPCO) for NRC appr9val of AFAT If data from the June 1996 Marble Hill annealing demonstration is needed to support AFATA, the proposed June 1996 approval date would be unrealisti Planriing efforts for performance of vessel annealing were thorough and in compliance with NRC Draft Regulatory Guide 102 Examples of good planning efforts include utilization of mockups to demonstrate feasibility and reduce exposure for placement of thermocouples within

. the annulus and selection and placement of pressure vessel materials in surveillance capsules, located in high neutron flux regions, to get additional information on the amount of recovery for annealed vessel materials. A key element is Palisades' participation in the Marble Hill annealing demonstration project. Data obtained during this effort will serve to validate thermal and stress analytical models to be used in*

Palisades' AFAT.4 Review of Information Notice (IN) 94-66. Supplement 1. "Overspeed Of Turbine-Driven Pumps Caused BY Binding In Stems Of Governor Valves" Th~ inspectors reviewed the applicability of IN 94-66, Supplement 1,

"Overspeed of Turbine-Driven Pumps Caused by Binding in Stems of Governor Valves," at the Palisades Plan The licensee has not had any experience with governor valve stem binding or corrosion problems associated with its turbine-driven auxiliary pum Pump components are inspected during routine preventive maintenance activities and pump performance is monitored during routine surveillance testin.5. Action On Previous Inspection Findings 3.5.l (Closed)

Violation 50-255/92-026-01: Lack of controls resulted in an operating procedure revision in conflict with system design requirement The appropriate procedures were revised to correct the specific problem and to address system operating limits in each system design basis document as described in Event Report (E-PAL)

93~001. The inspector reviewed the closed out E-PAL and the referenced procedure The change~ made were appropriate to address this issu.5.2 (Closed)

Violation 50-255/93-013-01: Activities affecting quality were not receiving required independent verification An event report, E-PAL-93-031, was initiated to revise Administrative procedure 5.2 The inspector ~eviewed Revision 2 of this procedure; Attachment 1 was revised to list the specific activities that required Independent Verification. A recent work order performing a pump alignment was also reviewed and it had the appropriate independent verification of the pump to motor alignmen.5.3 (Closed)

Violation 50-255/94-014-54: Allowing the condensate storage tank tempetature to exceed 120 degrees This violation did not require a response as the corrective actions had been reviewed as discussed in inspection report 255/9401.5.4 (Open) IFI (50-255/94014-64): Vendor recommendations for EOG not fully evaluated by Systems Engineerin The inspectors reviewed Corrective Action C-PAL-94-0728 which evaluated the diesel ~enerator vendor recommended maintenance practices against the licensee's current practice J~stifications for deviations from the vendor recommendations were include Presently the nuclear utilities in the ALCO.Owners Group are working to develop a maintenance program for their standby diesel generators based on their cumulative experienc The group includes a v~ndor representative. The vendor should concur with the developed program, once it is established, if it is not in accordance with the vendor's maintenance manual recommendations, in order for this issue to be resolve *

• PLANT SUPPORT NRC Inspection Procedures 71750 and 83750 were used to perform an inspection of plant support activities. The findings showed performance was adequat.1 Opening in Vital Area Boundary The inspectors identified that the main steam line penetration through

.the turbine building/auxiliary building wall contained an opening that was greater than the allowable size for an opening in a vital area barrier. A security officer was posted at the opening as an immediate compensatory measure and a condition report was initiated to document the finding and corrective action. Subsequently, the licensee determined that the penetration was not a viable pathway and relieved the security w~tc *

5.. 0 PERSONS CONTACTED AND MANAGEMENT MEETINGS The inspectors contacted various licensee operations, maintenance, engineering, and plant support personnel throughout the inspection perio Senior p~rsonnel are listed belo At the conclusion of the inspection on August 21, 1995, the inspectors met with licensee representatives (denoted by*) and summarized the scope and findings of the inspection activities. The licensee did not identify any of the documents or processes reviewed by the inspectors are proprietar R. A. Fenech, Vice President, Nuclear Operations

• T. J. Palmisano, Plant General Manager

• K. P. Powers, Engineering and Modifications Manager R. M. Swanson, Director, NPAD

• D. W. Rogers, Operations Manager

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D. P. Fadel, Engineering Programs Manager

• J. P. Pomaranski, Deputy Maintenance Manag~r

• H. L. Linsinbigler, Project Management and Modifications Manager S. Y. Wawro, Planning Manager K. M. Haas, Safety & Licensing Manager

• R. B. Kasper, Maintenance Manager

• C. R. Ritt, Administrative Manager R. M. Rice, System Engineering Manager 15