IR 05000255/1992023

Insp Rept 50-255/92-23 on 921006-1116.Noncited Violations or Deviations Noted.Major Areas Inspected:Operational Safety Verification,Reactor Trip,Maint Surveillance,Quarterly Mgt Meeting & Dry Cask Storage Project Insps

ML20128B698
Person / Time
Site:	Palisades
Issue date:	11/25/1992
From:	Jorgensen H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML18058B251	List: IR 05000255/1992023
References
50-255-92-23, NUDOCS 9212040055
Download: ML20128B698 (16)

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U. S. NUCLEAR REGULA10RY COMMISSION REGION 111 Report No. 50-255/92023(DRP)

Docket No. 50-255 License No. DPR-20 Licensee: Consumers Power Company 212 West Michigan Avenue Jackson, MI 49201 facility Name: Palisades Nuclear Generating Plant inspection At: Palisades Site, Covert, M1 Inspection Conducted: October 6 through November 16, 1992 Inspectors: J. K. Heller D. G. Pasuhl M. K. Gamberoni Approved By: - 3 7h __lk26c97-B? or ensen, GTcT Date Reactor Projects Section 2A insp_ection _ Summary inspection from October 6 throuah November 16. 1992 (Renort No. 50-255/92023(DRP Areas Inspected: Routine unannounced inspection by the resident inspectors of actions on previously identified items, operational safety verification, reactor trip, maintenance, surveillance, quarterly management meeting, and dry cask storage project inspections. No Safety Issues Management System (SIMS)

items were reviewe Results: No cited violations or deviations were identified in any of the seven areas inspected. One-non-cited violation (NCV) was identified pertaining to i control of an air supply valve to a service water valve (paragraph 3.b, -

"Hisaligned Air Supply Valve"). s The strengths, weaknesses, and non-cited violation are discussed in paragraph 1, " Management Interview." In summary, strengths were identified in rest,1ution of equipment problems and a weakness was noted in the number

, of plant trips during the current SALP cycl PDR 0 ADOCK 05000255 PDR j

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DETAILS 1. Manaaement Interview (71707) ,

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The inspectors met with licensee representatives - denoted in paragraph 9 - on November 20, 1992, to discuss the scope and findings of the inspection. In addition, the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspection was also discussed. The licensee did ,

not identify any such documents or processes as proprietar Highlights of the management interview are discussed below: Strengths noted: The integrity of the primary coolant system (paragraph " Containment Tour.") Troubleshooting and restoration activities for the "A" bus i failure to fast transfer, reactor trip and repair of a flow controller in the auxiliary feedwater system (paragraphs 4.b. --"A" Bus Fast Transfer failure", " Loss of Load", and " Maintenance").

The licensee stated that the root cause for failure of the

"A" bus to fast transfer may be applicable to other 4160 or 2400 volt circuit breakers. Work orders have been written to perform additional inspection . The thorough and competent manner in which three surveillance tests were performed (paragraph " Surveillance"). The strong working relationship between the site engineering staff and the plant management (paragraph 4.c. " Loss of Load"). Weaknesses noted: A third reactor trip was generated from the turbine control system. The total of reactor trips-for this SALP cycle is five (paragiaph 4.c. " Loss of Load").

- Work order documentation problem pertaining to post maintenance testing-(paragraph 5.a.." Maintenance"). A component was removed from service without a Technical Specification operability review (paragraph 3.b, " Misaligned Air Supply Valve").

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. The non-cited violation (NCV) pertaining to control of an air supply valve to a service water valve was discussed - (paragraph 3.b, "Hisaligned Air Supply Valve"). Seve al procedure enhancements were discussed (parsgraphs " Postulated loss of Charging Capability" and 3.d. " Containment Tour"). The need to be proactive when planning activities that may increase the frequency of a plant trip was discussed (paragraph " Maintenance").

Items f. and g. below were discussed at the meeting, but they had not been inspected and are not documer.ted in the repor The NRC Performance Indicators for Palisades plant were discussed, along with potential NRC actions for declining trend The licensee acknowledged the comment and stated that they are currently reviewing the latest published Performance Indicators to determine if improvements can be made in communications with the NRC concerning this information, The developing NRC position pertaining to changing electronic I components from analog to digital was discussed, along with potential impact on past modification .

The licensee acknowledged the inspector comment and stated that the general topic is being discussed with the NRR project manage . Acli_ons on Previously identified items (92701, 92702) (Closed) Open Item 255/90012-03(DRP): An uncontrolled copy of an instruction for installation of a solenoid valve was used during a maintenance activit The instruction did not have the documented reviews required by administrative procedures for control of vendor supplied information. The licensee determined that the instruction should have been considered a vendor manual and controlled by the veador manual control program. The instruction was included in the vendor manual control program, (Closed) Violation 255/90012-04(DPP): Several examples of a violation of the licensee's Fire Protection Implementing Program (FPIP) were note The licensee's response of June 8,1990, acknowledged-that the items were examples of noncompliance with the requirements stated in the FPIP. On the other hand, the licensee complied with the

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National fire Protection Association Standard Number 30 (NFPA-30),

" Flammable and Combustion Ligaid Code," a standard used to develop the FPI The licensee revised the FPIP to provide more flexibility in handling flammable liquids while remaining in compliance with NFPA-30. Additionally, the licensee conducted training for contractor and utility personnel to assure that the FPIP was understoo (Clored) Open item 255/90015-05(DRP): A piece of shrink tubing was used to prevent actuation of contacts during a post modification test. When the contacts were returned to service, there was no documentation that the shrink tubing was removed and that the contacts were not damage The licensee verified that the shrink tubing was properly remove During interviews with the personnel who performed the activity, the licensee verified that a visual inspection was performed to assure that the contacts were not damaged. The contacts functioned properly during subsequent surveillance testin Also, the licensee revised Engineering Guide EGAD-PROJ-12

" Guideline for Preparation of Modification Test Procedures" to recognize the use of shrink tubing and restoration verificatio (Closed) Violation 255/90018-04(DRP): A Technical Specification change was not requested when the feedwater control system was modified to include a feedwater isolation on a containment high pressure signal. The safety evaluation did not identify the need for a Technical Specification chang The licensee's response dated October 15, 1990, acknowledged the violation. They stated that their administrative procedures did not address a situation where a modification resulted in the need for added Technical Specifications. The administrative procedures have been revised to include prompts to assure that the evaluator performs the required raiews. Also, training was provided to the staff that performs safety evaluations to assure consideration of the need for added Technical Specifications, (Closed) Open item 255/90039-04(DRP): The monthly emergency diesel generator surveillance procedure alternated diesel starts with one of the two air start motors valved out of service. The test did not test the diesel in the normal condition with both air start motors in servic The licensee evaluated the open item and agreed that the monthly test should periodica11v include a start with both air start

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motors valved into service. Changes to the monthly surveillance procedure were mad _ _ _ _ _ _ - _ _ - _ _

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No violations, deviations, unresolved or inspector followup items were identifie . Operational Safety Ver_iLi.c11D n (71707, 71710, 42700)

Routine fecility operating activities were observed in the plant and from the main control room. Plant startup, post trip recovery, and steady power operation were observed as applicabl The performance of reactor operators and senior reactor operators, shift l engineers, and auxiliary equipment operators was observed and evaluate Included in the review were procedure use and adherence, records and logs, communications, shift / duty turnover, and the degree of profer;ionalism of control room activitie ,

Evaluation, corrective action, and response for off normal conditions were examine This included compliance to reporting requirement Observations of the control room monitors, indicators, and recorders were made to verify the operability of emergency systems, radiation monitoring systems, and nuclear reactor protection systems. . Reviews of '

surveillance, equipment condition, and tagout logs were conducte Proper return to service of selected components was verifie Gentrjd Tne plant began and ended the reporting period at essentially full

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power. A reactor trip occurred on October 30, 1992, which removed the plant from service for atu t one week. The trip is described in paragraph 4, " Reactor Trip." Misalioned Air Sunoly Valve During troubleshooting of the closed position indication for CV-0880 (ESS Pump seal cooling service water supply valve), a licensee electrician found its air supply val'.a close Valve CV-0880 is an air-operated normally closed valve installed in one train of the critical service water heade It functions to supply backup cooling water to the Engineered Safeguards Pump Seals. The normal seal cooling supply is from the Component Cooling Water Syste Approximately 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br /> earlier, Operators had completed satisfactory open stroke time testing of CV-0880 using surveillance test procedure QO-5, " Valve Test Procedure (Includes Containment Isolation Valves). Upon re-closing CV-0880, a licensed auxiliary operator (LAO) noted the closed indicating light had not energized. The LA0 initiated a Work Request to correct the problem. in order to complete the " Problem Description" section of the Work Request form, the LA0 sought to determine whether the problem was due to a mispositioned limit switch or air leakage past the air supply solenoid valve keeping

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CV-0880 off its fully closed seat. The LA0 unilaterally isolated the air supply to check for air leakage without realizing he rendered CV-0880 INOPERABLE. After concluding that the limit switch was mispositioned, the A0 failed to return the air supply to servic The inspector reviewed Technical Specification (TS) 3.4,

" Containment Cooling," and verified that the applicable TS was met, albeit accidently. Safety significance was low since seal cooling to any of the ESF pumps can be supplied by the common CCW header or either train of Service Water. Af. the time CV-0880 was INOPERABLE, there were two redundant sources of seal cooling available to the ESF pumps. Also, the seals are designed for temperatures around 300 degrees F., and it is extremely unlikely the seals would see this condition. Finally, the FSAR implied seal cooling was not required for pump operability since cooling is only provided to extend seal lif More important was whether the licensee's procedures and training allowed Operator re)ositioning of safety related equipment in the plant without a Wor ( Order er knowledge of the Shift Superviso The inspector did not find specific instructions that require shift supervisor approval prior to changing the position of safety-related equipment, nor is there a requirement to review the 1S to ensure Action Statement compliance. However, a review of step 5.2.2 of Administrative Procedure 4.0, " Control of Equipment Status," found that the final position of manual valves, including those valves in the Critical Service Water Header, require final position verification by a repositioner and an independent verifier. The failure to verify the position of the manual air valve for CV-0880 is considered a violation of this procedure and thus a violation of TS 6.8.1.a. However, the licensee identified this violation, and it is not being cited because the criteria specified in Section VII.B.2 of the " General Statement of Policy and Procedures for NRC Enforcement Actions," (Enforcement Policy, 10 CFR Part 2, Appendix C (1992)), were satisfied. The licensee's corrective actions included a memo issued to operators discussing this incident and emphasizing management's expectation that shift supervisor permission be obtained prior to altering the position of equipment not covered by procedures or that are not part of normal operator dutie This item was discussed at the management interview. The inspector noted that preparation of a memo was a good interim solution. However, a change to the administrative procedures to include the essence of the memo may be a more permanent enhancement to the program, c. Postulated Loss of Charaina Capability The inspector met with the operations superintendent to discuss loss of all charging capabilit The discussion arose when,

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I during power operation, two of three charging pumps became inoperable. In this instance, Charging Pump No. P-55A was  ;

inoperable for pre-planned maintenance, while P-558 unexpectedly '

became inoperable due to a low oil pressure condition. The remaining pump, P-550, remained in servic The inspector questioned the control room operators on actions they would take subsequent to loss of P-550. No saecific guidance existed for operators on what actions to take if t1e third pump was inadvertently lost. Although loss of all charging is somewhat addressed in the 10CFR50 Appendix R Alternate Safe Shutdown Procedure (OHP 25.2) and several emergency operating procedures, the operators had no clear instructions to lead them from power operation into those procedures. The existing alarm response procedures addressed a degrading charging capability but assumed at least one charging pump was availabl The operations superintendent and shift supervisor discussed the situation at the onset of this event and concluded that, upon failure to return at least one cht.rging pump to service, they would direct operators to trip the reactor and enter the emergency o)erating procedures. To ensure clear guidance exists to the s11fts in the future, the operations superintendent indicated that a revision to the alarm response procedures addressing this matter would be forthcoming, d. Containment Tour

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The inspector accompanied licensee )ersonnel on a routine tour of the containment building. During tie tour, an auxiliary operator performed a routine measurement of the control rod drive mechanism leak rate, which showed no advcrse trend. A line item of the general checklist in GOP 3, "flot Shutdown to Critical in llot Standby," was satisfactorily performed to check for pressurizer head vent leakage. The containment building was generally clean, with or.ly one notable boric acid leak on a discharge valve (M0V-3041) from one of the safety injection tanks. The licensee was aware of the leak and had previously written a work request to repair the valve during the next refueling outag Upon exiting the containment building, the inspector observed a fire extinguisher that hadn't been checked since April 1992. The inspector noted the same was true for all extinguishers inside the containment building after reviewing checklist No. 21.5, " Fire Extinguisher Checklist - Auxiliary Building." The checklist stated that the fire extinguishers in containment '...only need to be checked during shutdown condition." The inspector questioned this statement, since the unit had been in a shutdown condition five times since July 1992. The licensee stated that, like the fire hose stations inside containment, the fire extinguishers are only inspected c e e per month during refueling or maintenance outages when the containment is accessible, if there was a fire

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inside the containment, the fire brigade would provide the proper respons The inspector agreed this was satisfactory since the five forced outages were neither maintenance nor refueling outages. The licensee agreed to change the language in checklist No. 21.5 so their practice and program are the sam One non-cited violation was identified. No deviations, unresolved, or inspector followup items were identifie . ReactorTrio(93702) Qqngul The reactor automatically tripped at 07:00 a.m. on October 30, 1992, due to a Reactor Protection System loss of Load signal generated from a Digital Electro-hydraulic (DEH) Control System Turbine Trip actuation. The NRC inspector responded to the control room to assess control room activities; no problems were noted with command and control of the event. The inspector interviewed control room operators, reviewed control room indications, and determined that all safety systems operated properly. However, the "A" non safety-related 4160 volt bus did not fast transfer to its alternate power supply. The unit remained in HOT SHUTDOWN for about one week to investigate the cause and make repairs. The unit was returned to service on November 8, 199 ."A" Bus fast Transf_cr failure The cause of the "A" bus fast transfer failure was a vibration induced actuation of the alternate supply breaker's foot / trip plunger. The plunger is used when the breaker is racked from its cubicle. The plunger is coupled to the breaker trip device as a aersonnel safety feature to prevent personnel from removing a live areaker from servic Troubleshooting activities determined that the breaker was misaligned within its cubicle making it susceptible to vibration and inadvertent actuations of the foot / trip plunger. This misalignment was initially identified and resolved during fiald installation of the breaker at the time the plant was constructed, but these field settings were not adequately documented. When the 4 breaker was refurbished during the last refueling outage, all f settings were returned to original factory settings and not the 3 corrected field settings, following this reactor trip, the correct settings were established and adequately documente Additionally, a locking device was installed to restrain the foot / trip plunger. The locking device does not prevent actuation of any automatic or manual (hand switch) trip. feature The "A" Bus had previously failed to fast transfer during the July 1, 1992, reactor trip. At that time, the licensee did not

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identify a specific root cause - that breaker was replaced with a spar c. Loss of Load The cause of the loss of load and subsequent reactor trip was never precisely determined. The licensee determined that the loss of load was generated from the DEH computers. This was the third loss of load trip generated from the DEH computers since July 1, 1992. The previous trips were discussed in Inspection Reports 255/92018(DRP) and 255/92022(DRP).

The most probable trip sequence involved degraded line voltage from a supply transformer that caused the DEH Uninterruptible Power Supply (UPS) to experience a momentary low input voltage and subsequent low output voltage. This caused the power supplies in the DEH computers to fault, tripping the turbine and then the reactor. The licensee retrieved the stored information from the UPS internal memory at the time of the DEH failure. This information confirmed the degraded input and output voltage condition. In addition, the licensee found that the incoming line voltage from the supply transformer was approximately 10 percent below its nominal voltage. This was corrected. The cable to the DEH room was also enlarged to minimize voltage losses on the line i tsel During restoration activities, the site NRC inspection staff, with assist 6nce from a Region III electrical specialist and the Palisades Region III Section Chief, observed troubleshooting and problem solving sessions conducted by an engineering group headed by a system engineering section chief. The group consisted of members from the site design and system engineering departments, electrical and maintenance personnel, and representatives from the vendors for the DEH and UPS. Plant management 3rovided the group with repair and restart acceptance criteria. 11e engineering group routinely briefed management on root cause analysis, repair options, schedules, and restoratio Several interim corrective actions were implemented. They included installation of new power supply modules in the DEH computers, a new UPS, SOLA trar.sformers supplying some of the DEH computers, separate power sources to the redundant DEH computers, and additional alarmt to alert the control room operators to a DEH power supply problem. These actions only provided redundancy in the power supply from Motor Control Center 14 to the DEH computers. Final changes to the DEH power supply configuration are being researched for implementation during a future outage, d. Mditional NRC Inspection Ltivitiet The site inspection st9ff performed several walkdowns of the electrical lineup from Oe 2400 volt "E" bus to the DEH computers.-

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The walkdowns were performed with members of the operations department to identify loads or breakers that could affect power to the DEH computers. The walkdowns identified several breakers '

that required labelling changes to properly identify their new function. Also, several breakers were identified that provided power to welding sources or outage trailers that could indirectly affect the )ower sup)1y to the EH computers. Caution tags were placed on t1ese brea(er This was the third loss of load reactor trip since the licensee installed the DEH during the 1992 refueling outage. The inspectors reviewed the modification )ackage to determine if the licensee evaluated and incorporated tie lessons learned from other utilities who experienced problems with similar DEH control systems. The licensee had contacted those other utilities and took action to correct and )revent similar problems at Palisade The problems associated wit 1 the recent reactor trip are new but may have been prevented with more rigorous post-modification testing such as monitoring the line voltage from the incoming transforme e. Plant Review Committee (PRC)

The NRC site inspection staff and the assigned Region III Section Chief attended t1e startup PRC meeting conducted on November 6, 1992 The items discussed above were presented to plant mana ement by knowledgeable individuals who described the prob ems, corrective actions, and the technical merit of the solution The inspectors verified that the PRC commition met technice'

specification manning requirements and 11at a voting quorum was presen f. Telephone Conference with the NRC Shortly after the trip, senior Region 111 management notified senior Consumers Power corporate management of the need for a conference call to discuss root cause and corrective and preventive actions prior to returning the plant to servic The conference call was held at the completion of the startup PRC meeting. The Itcensee's site managers conducted this conference call with the NRR Project Manager, NRR Section Chief, Region Ill Branch Chiefs (DRP and DRS), Region 111 Section Chiefs (DRP and DRS), and the site inspection staff. During this conference call, the licensee discussed the cause of the trip, root cause analysis, and interim corrective actions, and they responded to several question _.__ _ _ . _ , _

. Return to Service The inspector observed the initial turbine roll and turbine trip testing arior to returning the unit to service on November 8, 199 Tae operations superintendent was the on-shift reactivity manager during these activities. The evolution proceeded as expected with no problems note No violations, deviations, unresolved or inspector followup items were identifie . Maintenance (62703, 42703)

Maintenance activities were routinely inspected, including both corrective maintenance (repairs) and preventive maintenanc Mechanical, electrical, and instrument and control group maintenance activities were included as availabl The focus of the inspection was to assure the maintenance activities reviewed were conducted in accordance with approved procedures and in conformance with Technical Specifications. T!.9 following items were considered during this review: the Limiting Conditions for Operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using aaproved procedures; and post mainter ace testing was '

performed as applica)l The following work order (WO) activities were inspected: WO 24204865, " Troubleshoot and Repair Auxiliary feedwater System flow Controllers FIC-0737A and FIC-0736A."

A common 26 Volt d.c. power supply to Auxiliary feedwater (AFW)

System Flow Instrument Controllers FIC-0737A and FIC-0736A had degraded to approximately 23.5 Volt d.c. as I&C technicians began a scheduled calibration procedure (RI-95D) on the flow controllers. A " power error" light energized on the controllers when the voltage degradation occurred. The controllers send position demand signals to Auxiliary feed Water (AFW) flow control valves CV-037A and CV-036 Both valves are at the discharge of AFW Pump P-8 Upon receipt of the " power error" indication, a Control Room Operator attempted to manually cycle CV-037A without succes The licensee immediately declared P-8C inoperable and entered a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Technical Specification Action Statement. The flow path from the two other AFW pumps remained operable throughout this tim Instrument and Control (l&C) personnel performed an excellent troubleshooting job and quickly found the cause to be a failed capacitor in the power supply to the controllers. The 1&C system

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i engineer's involvement in and knowledge of the control system wer7 also excellent. A new qualified capacitor wa:. installed upon removal of the failed capacitor. Post-maintenance testing of the power supply was satisfactorily performed. The AFW Pump was returned to operable status within the required time frame. The 1&C staff also did an excellent job of documenting work performe On the other hand, documentation of post-maintenance testing of the AFW flow control valves was weak. The computerized Work Order did not capture any return to service testing of the valve Ai-o, the original Work Order did not describe the location where th valve stroking was performed and whether the results were r isfactory. Since the controllers do not provide actual valve position but only demanded valve position, an operator should have been dispatched to locally observe valve response and this should have been stated. The inspector questioned the operator who initialed the Work Order, and he stated that an auxiliary operator was dispatched to the four valves to locally verify they operated correctl t b. WO 24204713, " Replace The Reactor Protective (RPS) "AC" Matrix Power Supply From (-30" During routine control room operator rounds, an unusual lighting configuration in the "AC" RPS matrix was observe Troubleshooting activities determined that one of two power supplies in the matrix had degraded and could fail. The system configuration was designed to permit operation with one failed power supply, but a failure of the second power supply would deactivate the "AC" matrix and cause a reactor tri The system engineering group determined that a replacement of the failed power supply was justified. However, the utility had no experience with on-line replacement of the power supply. Prior to the replacement, confirmation was obtained from the vendor that the system was designed for on-line replacement. Additionally, contact was made with other utilities to assure that the replacement had been and could be satisfactorily performed on-lin The inspector attended a management meeting where system engineering presented and discussed the options with site management. The power supply was successfully replaced on October 6, 199 c. WO 24204995, " Replace DEH and UPS Power Supply Per Facility Change (EDC-FC-844-27)."

d. WO 24205061, " Install Sola Transformers, Conduit and Cables (EDG-FC-844-29)."

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No violations, deviations, unresolved or inspector followup items were identifie . htyg.111ance (61726, 42700)

The inspector reviewed Technical Specifications required surveillance testing as described below and verified that testing was performed in accordance with adequate procedures. Additionally, test instrumentation was calibrated, Limiting Conditions for Operation were met, removal and restoration of the affected components were properly accomplished, and test results conformed with Technical Specifications and proce.htre requirements. The results were reviewed by personnel other than the

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individual directing the test and deficiencies identified during the testing were properly reviewed and resolved by appropriate management ,

personne The following activities were inspected: " Inservice Test Procedure - Charging Pumps" "

Safety injection System" The previous performance of QO-1 on J 26, 1992, resulted in a plant voltage transient that faulted , . AH computers and caused a reactor trip (Ins >ection Report 255/95 22(DRP)). The corrective actions following t1at reactor trip included installation of a UPS to minimize the effect of similar voltage transients. Tne UPS  !

performance was satisfactory on October 26, 1992, with no adverse effect to the OEH computers,

• Hi-1 Nuclear Power Rarge Rod Drop Alarm Flux 'lest" MI-2 "

Thermal Margin and Low Pressure Trip" The inspector observed portions of Q0-1, Mi-1, and Mi-2. The tests were accomplished by different work groups and performed on different day The inspector noted a thorough discussion pertaining to the purpose of the tests, the expected results, and the actual results. Additionally, there was independent verification that the item to be manipulated was positively identifie No violations, deviations, unresolved or inspector followup item.; were identifie . . Quarterly Manaaement Meetino (30702)

A Quarterly Managemcot meeting was held at the Region Ill Office on October 30, 1992, with the personnel indicated in paragraph 9 in attendanc p,aw-amu-- --w-a--= cec %-mcvg- r *m- y 9C r-- y---+ y- ye-- v-v yy---'m---,w y- ---g- y, e- y-tw "w -'7%'? -

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ihe following items were discussv.:

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A summary of the recent reactor trips that occurred on July 1, July 24, August 14, and August 24, 1992. The discussion focused on causes found and corrective actions implemente A review of the September 18, 1992, electrical fatalit A discussion of the September 20, 1992, turbine generator electro-hydraulic control problem An overview of the Spent fuel Storage Project including work remaining and time line sequence for cask loading and movemen An assessm?t of experience and effectiveness of the Quality Verification Progra A review of the NRC inspection program at Palisade No violations, deviations, unresolved or inspector followup items were identifie . Dry Cask Storace Proiect Inspection (42700, 62703)

.I Tt.e inspector reviewed the handling procedures, crane inspections, and operator training related to the Palisades Dr/ Cask Storage ProjecL These areas were found to be satisfactory, Procedures Per NUREG 0612, " Control of Heavy Loads at Nuclear Power Plants,"

procedures for handling heavy loads should include identification of requi_ red equipment, inspections and acceptance criteria required before movement of load, the steps and proper sequence to be followed in handl'ng the load, and definition of the safe load path. Procedure FHS-N-23," Movement of Heavy Loads in the Spent fuel Pool Area," adequately met the guidelines in NUREG 0612. In addition, Palisades is currently preparing specific procedures for load testing the spent fuel pool crane and loading and unloading of the dry casks. The loading and unloading procedures which are expected to have specific sequencing, will be reviewed as part of l ongoing inspection of this project, Inspection

" Overhead and Gantry Cranes," Chapter 2-2, " Inspection, Testing, and Maintenance" (ANSI B30.2), includes daily, mLnthly, and annual inspection directions and data sheets which document ANSI requirement Procedure MSM-M-13, " Overhead Crane Mechanical Inspection," met the requirements of ANSI B30.2. Furthermore, MSM-M-13 also contained a work request for the annual dye penetrant inspection testing of the L-3 auxiliary and main hook;

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L however, it incorrectly referenced the L-2 auxiliary and main hook. The licensee sgreed to change the procedure to reflect the correct referenc A condensed listing of the work order history on the crane for the past three years was reviewed. The list contained all monthly and yearly electrical and mechanir.al preventive r'ntenance. In addition, corrective maintenance, rework, an- .lergent work were also listed. No probi 3ms were note Trainina Per NUREG 0621, crane operators should be trained in accordance with ANSI B30.2, " Overhead and Gantry C ' .." Chapter 2-3,

" Qualifications for Operators." Chaptr requires operators to be familiar with operator conduct, hare .., of loads,. equipment inspection, and standard sigi'is. Palisades has a lesson plan for overhead and gantry crane cperators' training. This training includes both classroom and hands-on instruction. The lesson plan included the ANSI B30.2 requirements listed above. Also, ;1ysical qualification such as good vision and hearing are required. A master list was revievied and showed all personnel qualifications as curren No violations, deviations, unresolved, or inspector followup items were identifie . Persons Contacted

[pg,imers fower Comoany

1. D. P. Hoffman, Vice President, Nuclear Operations

• 1. G B. Slade, Plant General Manager

• 1. T. J. Palmisano, Plant Operations Manager a D. Orosz, Nuclear Engineering and Construction Manager

1. R. M. Rice, Director, Nuclear Performance Assessment Da; Went

• 1. P. 3. Or ally, Safety and Licensin5 Director K. h. 6 4, hdiological Services Manager J. L. Hansoc, Operations Superintendent

1. R. B. Kasper, Maintenance Manager

• K. E. Osboiie, System Engineering Manager D. G. Malone, Operations Staff Surpart Supervisor

• W. L. Rober 1, Senior Licensing Engineer R. W. Smedley, Staff Licensin, Engineer

1. K. A. Toner, Electrical / Inst: nt and Control / Computer Enginearing Manager

1. L. J. Rc Maintenance Superintendent

1. T. Jones. Electrical Maintenance Technician

1. R. Dixon, istrument and Control Technician

1. L. T. Tay. , Mechanical Maintenance Technician fJ. T. Hova. Senior Electrical Engineer

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1. M. A. Ferens, Project Manager, Spent fuel St'orage D. Engle, Project Engineer, Dry Cask Storage Nuclear Reaulatory Commission (NRC)

i #H. J. Miller, Director, Division of Reactor Safety

. #C. J. Paperiello, Deputy Regional Administrator

1. W. L. Forney, Deputy Director Division of Reactor Projects

1. W. D. Shafer, Chief, Reactor Projects Branch 2

• 1. B. L. Jorgensen, Chief, Reactor Projects Section 2A

1. W. G. Snell, Chief Emergency Preparedness and Effluents Section

• J. K. Heller, Senior Resident Inspector

• 1. D. Passehl, Resident inspector

1. Derotes some of those present at the quarterly management meeting on Ori vber 30, 199 * Denotes some of those present at the exit meeting on November 20, 1992.

Other members of the plant staff, and several members of the contract security force, were also contacted during the inspection period.

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