IR 05000155/1999003

Insp Rept 50-155/99-03 on 990416-0608.No Violations Noted. Major Areas Inspected:Facility Mgt & Control,Decommissioning Support Activities,Spent Fuel Safety & Radiological Safety

ML20196E188
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	06/21/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20196E163	List: IR 05000155/1999003 ML20196E160
References
50-155-99-03, 50-155-99-3, NUDOCS 9906280102
Download: ML20196E188 (20)

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U.S. NUCLEAR REGULATORY COMMISSION REGION 111 Dockst No: 50-155 License No: DPR-06

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Report No: 50-155/99003(DNMS)

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Licensee: Consumers Energy Company Facility: Big Rock Point Nuclear Power Plant Location: 10269 U.S. 31 North Charlevoix, MI 49720 Dates: April 16,1999 - June 8,1999 Inspectors: R. J. Leemon, Reactor Decommissioning Inspector l R. B. Landsman, Project Engineer l P. W. Harris, License Project Manager W.G. Snell, Health Physics Manager M. M. Lafranzo, Radiation Specialist Approved By: Bruce L. Jorgensen, Chie l Decommissioning Branch Division of Nuclear Materials Safety 9906290102 990621 PDR ADOCK 05000155 G PM

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EXECUTIVE SUMMARY Big Rock Point Restoration Project NRC Inspection Report 50-155/99003(DNMS)

This routine decommissioning inspection covered aspects of facility management and control, decommissioning support activities, spent fuel safety, and radiological safet * Overall, major decommissioning activities were properly controlled and were performed in accordance with schedules. The licensee effectively monitored both the conduct and the cost of decommissioning. The second phase of the spent fuel pool clean out project, which involves removal of significant quantities of radioactive material, is well underway. Three shipping casks, each loaded with 13 crushed control blades, were successfully loaded and shipped to a licensed burial site. The minimum operations shift was reduced to a complement of two person Facility Manaaement and Control e Nuclear Performance Assessment Department appeared to be effectively performing their function; the decommissioning audit identified several conditions adverse to qualit The corrective action system was functioning as designe * The material integrity of systems, structures and components necessary for the safe storage of spent fuel and conduct of safe decommissioning was being maintained, housekeeping and control of combustible materials were good, and fire equipment was being properly maintaine Decommissionina Suocort Activities

• Daily work activities were performed well on the reactor deck. When problems were encountered, the supervisors were notified and the activity was stopped, the correct contingency plan was generated and executed or repairs were made to processing equipment, and condition reports were written so the problems could be evaluated by managemen * Good worker performance was observed during the second phase of the spent fuel pool clean out project. NRC and Department of Transportation requirements were me * On June 7,1999, the operations department established the final shift staff complemen The operations shifts are now staffed with one shift supervisor who is a certified fuel handler, and one decommissioning worker class 1. This constitutes the two person shift authorized by the Defueled Technical Specifications for non-fuel handling condition Soent Fue! Safety -

• The safety of spent fuel was being monitored and maintained. On May 5,1999, the new spent fuel pool cooling system was declared operabl .

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Radiolouical Safety -

• As-Low-As-Reasonably-Achievable and radiation work permit briefings forjob tasks were being performed well. Remote TV monitoring of spent fuel pool activities was provide e items noted as meriting licensee attention included: 1) adequacy of knowledge of monitoring station operators regarding emergency response information to be provided if an incident should occur during transport of radioactive material; 2) use of a defective survey instrument while performing a radiological survey; and,3) erroneous use of a turque multiplier, which resulted in the under-torque of the Fort Saint Vrain (FSV-1) cask bolts. The licensee planned appropriate actions for each ite e The loading of a shipping cask onto a transport carrier and its subsequent shipment to a licensed burial site were observed to be carried out as required, with associated documentation and record keeping completed as specified by procedure. No problems were identified with the activities surrounding the shipmen * Radiological control activities in the areas of effluents and solid radwaste were being conducted safely and in compliance with applicable requirement ,

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Report Details Summarv of Plant Activities During the inspection period, equipment not necessary for the safe storage of spent fuel and potentially hazardous components and materials were removed from the facility. Also, additional plant systems were classified as not needed for safe storage of fuel. The second phase of the spent fuel pool (SFP) clean out project, involving removal of sigr*icant quantities of radioactive materials, ns well under wa .0 Facility Management and Control General The inspector conducted frequent reviews of ongoing plant activities and attended licensee meetings and reviews addressing these activities, in order to assess overall facility management and controls. Specific events and findings are detailed in the sections belo .2 Oraanization. Manaaement. and Cost Controls at Permanent!v Shut Down Reactors (36801)

The inspector selectively reviewed the licensee's activities involving overall management and control of the decommissioning process. The effectiveness of the licensee's review of regulatory information applicable to the facility was selectively examine .3 Safety Reviews. Desian Chanaes. and Modifications at Permanently Shut Down Reactors (37801) Inspection Scoce (42700)

The inspector monitored the licensee's safety review program to determine if the program was effective at identifying potential unreviewed safety questions (URSQs) in accordance with 10 CFR 50.59. The activities of the licensee's onsite safety review committees (SRC) were evaluated to determine whether they were fulfilling their respective chcriers, the requirements of Derueled Technical Specification (DTS) and the '

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requirements of the quality assurance (QA) pla Observations and Findinas l

When attending SRC meetings, the inspector observed that a quorum was present and that the members understood the change to the procedure or plant modification under discussion. The members held challenging discussions on items and at times roerned procedures or modifications to the author for enhancements prior to determining that the change did not create an URS .

SRC members discussed and reviewed procedure changes to ensure they were clear, accurate, and had the same meaning to all members of 'he committee. Proposed procedure changes related to the processing of 10 CFR 50.59 and 10 CFR 50.82 evaluations were retumed to the author for enhancements. These proposed changes will be retumed to the SRC at a later date for approval. Human factors were considered

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In the changing of these procedures. Each procedure change was evaluated to determine if an 'URSQ was created. None were create Conclusion Based on these observations, safety committee discussions were challenging and there l were no issues which required changes to Technical Specifications (TSs), no URSQs were identified, and the requirements of 10 CFR 50.82 were me ' Monitored Decommissionina Activities The inspector attended licensee meetings where the planning, reviewing, assessing, and scheduling of decommissioning activities were observed. The inspector ascertained that activities we re in accordance with licensed requirements and docketed ]

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commitments as stated in 10 CFR, DTS, Updated Final Hazards Summary Report (FHSR), Post-Shutdown Decommissioning Activities Report (PSDAR), and Defueled Emergency Plan (DEP). Decommissioning activities monitored by the inspector were as described in the following section . Plant Systems and Components Removed from Service Pint systems and components detennined not needed for the safe storage of opent )

fuel and permanently removed from service during this inspection period were: *

- portions of the control rod drive system

. portions of the main condenser e portions of the post incident shutdown cooling system 1.4.2 Plant Modifications Specific design changes or modifications were reviewed to assess program effectiveness. This included a review of written safety evaluations and ottar records. In addition, a sample of maintenance and repair activities was reviewed to ascertain whether the licensee had made changes to the facility without properly invoking their safety review proces Plant mods. cations that were in progress or completed during the inspection period j were:

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planning for dry cask storage of spent fuel 1

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completion of the addition of plates to the new SFP heat exchangers

- completion of the SFP cooling system modification

- execution of the SFP clean out project Phase 2 Two major criteria used by the licensee for scoping decommissioning activities were

' annual budgeted money and budgeted person-rem exposure. No concems were identified with the scope and status of monitored decommissioning activitie .

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1.4.3 Manaaed Plant Hazard Reductions l

The primary hazardous materials removed from the plant during the inspection period were asbestos, which is a biological hazard. Approximately 88 percent of the 1 asbestos on plant systems has been remove i Self-Assessment. Auditina. and Corrective Action (40801) j 1. General The licensee's controls for identifying, resolving and preventing issues that degrade safety or quality were examined, including self-assessments, auditing, corrective i actions, SRCs, root cause eva!uations and third partv reviews. The inspector also 1

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attended corrective action review boards (CARBs) and management review boards (MRBs).

The QA plan and applicable implementing procedures forrned the basis for the inspection. In addition, the procedures were evaluated from the perspective of their adequacy to accomplish the objective of assuring that management and staff are knowledgeable of plant / activity performance and contribute effectively to safety and quality in conduct of irr,wrtant activitie .5.2 Independent Assessment of Bla Rock Point Activities Insoection Scooe (40801)

The inspector held discussions with Nuclear Performance Assessment Department (NPAD) assessors, attended NPAD exit meetings, and reviewed NPAD audits and reports and condition reports written by NPAD. The condition report provides tracking of the finding and assigns resources to determine the root cause of the finding so that the root cause can be corrected, Observations and Findinas The inspector observed NPAD actively monitoring decommissioning activities and pre-job briefings, and providing management timely feedback on their concems. The onsite leader of NPAD met with plant management at a planned monthly meeting to discuss NPAD's findings, and to answer any questions on the findings. Conditions identified by NPAD as adverse to quality were entered into the plant corrective action system to ensure followup. A ten member NPAD team performed the third decommissioning plant all department audit. The audit team identified no ,

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significant conditions adverse to quality, however; the audit team did identify ten minor i compliance conditions adverse to quality. These conditions were discussed with the plant management at the audit exit. Managen. ant agreed with the designation of these conditions as adverse to quality. NPAD wrote condition reports for the conditions, which l entered them into the plant corrective system for followup. NPAD appeared to be l l

effectively performing its missio l When the licensee discovered concems or abnormal conditions, the appropriate ccndition report was written and d4 postioned by the CARB. The inspector reviewed condition reports, closed corrective actions for condition reports, and reviewed trend

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reports for the corrective action system. The inspector also attended CARBs and MRBs. The corrective action system was being managed, generally correct root causes were being identified, and corrective actions were being closed in a reasonable tim . Conclusion NPAD appeared to be effectively performing their function and the decommissioning audit identified several conditions adverse to quality. The corrective action system was functioning as designe .6 Decommissionina Performance and Status Review at Permanentiv Shut Down Reactors (718011 1. General The performance of the licensee and contracted workforce in performance of

decommissioning activities was evaluated to determine if these activities were performed in accordance with licensed requirements and commitments. Control and conduct of facility decommissioning activities were examined to verify the requirements of the license and DTS were followed, and to ensure the requirements and commitments described in the Updated Final Hazard Summary Report, PSDAR, and DEP were followed. Specific events and findings are detailed in the section belo I 1.6.2 Plant Tours to Evaluate Material Conditions. Housekeeoina. and F!re Protection I inspection Scope (71801)  :)

'i The inspector conducted plant tours to evaluate the material integrity of

. systems / structures / components (SSCs) necessary for the safe storage of spent fuel and conduct of safe decommissioning, to observe and assess the status of facility housekeeping, and to evaluate fire protection issues such as control of combustibles, condition of fire fighting equipment and presence o.".he plant fire brigad Observations and Findings Observations from plant tours showed that the material integrity of SSCs important to safe storage of spent fuel (ISSSF) was being maintained. The inspector discussed tour observations with plant management. The inspector also observed that plant management was actively monitoring plant material condition Housekeeping observations focused on the areas adjacent to and containing SSCs necessary for the safe storage of spent fuel, effluent control, or radiation protection and rnonitoring. The inspector determined that all areas of the plant were kept clean and free of the accumulation of dismantiement debris. As material was disassembled, it was ,

' placed into metal boxes. ' Portable cables were routed so as not to cause tripping j hazards., Pre-job briefings for work on the reactor deck contained discussions of l i

possible tripping hazards because of the wire and hoses spread across the reactor deck. The reactor deck crew removed all unnecessary hoses and cabies and carefully i routed the others to reduce tripping hazards. General area housekeeping was good

with no areas of inspector concem during this perio l 7 i

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The storage of combustible and flammable materials were within the fire loading limits for transient combustibles and combustible materials were not accumulating in the plan The fire fighting equipment and stations were properly maintained and ready for us The installed fire detection and suppression systems were effectively maintained, surveillanco tests were performed, and the equipment was capable of performing the intended function Active dismantlement activities which included cutting and grinding were performed with the proper fire protection in place. No fires were started during these activitie Conclusierl

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Material integrity of SSC's important to safe storage of spent fuel and safety in decommissioning was being maintained. Housekeeping, control of combustible materials and operation of fire equipment were properly maintaine .7 Onsite Followuo. Written Reoorts of Non-routine Events at Power Reactor Facilities (92701)

i 1. (Closed) VIO 50-155/98004-02: Improper use of and inadequate Reliability Criteria:

The measure used in the demonstration for system functions under the NRC

• Maintenance Rule" consisted only of an inappropriate reliability measure, one which allowed repetitive maintenance preventable functional failures that are a measure of corrective actions rather than reliability. As a result, it was not demonstrated that effective preventive maintenance ensured that systems remained capable c, performing as require . The licensee's letter dated July 8,1998, responding to the Notice of Violation, established new allowable repetitive maintenance preventat'le functional failure criteria for the systems in question. The facility is currently in full compliance, achieved by the revision of performance criteria and the completio.1 of the review of system function This violation is close .7.2 (Closed) IFl 50-155/98002-05(DNMS): On February 24,1998, a worker at the GTS/Duratek waste processing facility in Oak Ridge, Tennessee, discovered metallic mercury leaking from a bag of waste that had been shipped from Big Rock Poin The preliminary root cause of this event was discussed in NRC Inspection Raport 50-155/98002(DNMS), and the licensee initiated a Condition Report, CR-BRP-98-0038, to address th's incident. Corrective actions by the licensee included establishing a program to sort 100 percent of waste, placing purple tape on all items comprising hazardous waste, and instituting independent checks during waste packaging for shipment to ensure no prohibited materials are loade l

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. l Decommissioning Support Activities Maintenance and Surveillance at Permanentiv Shut Down Reactors (62801) j The inspector evaluated maintenance and surveillance on SSCs potentially affecting the safe ctorage of spent fuel and reliable operation of radiation monitoring and effluent I control equipment. Direct observations, reviews, and interviews of licensee personnel were conducted to evaluate the proper implementation of DTS, and 10 CFR 50, Appendix B requirement .2 Dismantiement Activities (62801) 1 Dismantlement activities observed or evaluated by the inspector during the period were:

a removal of portions of the control room

- removal of control rod drive system piping

. removal of equipment from the electrical sub station a removal of equipment from the electrical and air compressor room

= removal of equipment from the containment Interior cable penetration room j

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. condenser dismantlement

- pipe tunnel dismantlement

• attemate shutdown building dismantiement The inspector observed that proper maintenance, personnel safety, fire protection, radiation protection, and health physics survey practices were performed during these activitie .3 Soent Fuel Pool Clean Out Proiect Activities j Inspection Scope (62801)

The SFP clean out project involves the removal of all norafuel bearing components from the SFP. For this project, the reactor vessel was being used as a water shielded processing tank. Each component was surveyed, radioactively characterized and

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classified, processed, packaged, and shipped for burial - inspection of radiological control activities is discussed in Paragraph 4.3 below. The inspector observed workers performing various of thess 1ctivities in and around the SFP and the reactor cavity. The inspector also attended planning meetings, attended pre-job and as-low-as-reasonably-achievable (ALARA) daily briefings and reviewed the SFP clean out project contingency plan b. Qb.Eprvations and Findinas The SFP project meetings were well organized. The daily plans were described, including possible " fill in" work identified at the previous day's planning meeting. Work experiences from the previous day were discussed at each meeting. Contingency measures for unexpected radiologMal conditions were provided, along with diccussion of ,

the overall ALARA plan for the project. Worker participation was fostered by asking questions and by generating interactions among the various work groups. Daily, staff were reminded that radiation protection (RP) had me authcrity to stop work at any time that the dose rate or P2 practices were in questio .

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Daily work activities were performed well on the reactor deck. When problems were encountered, the supervisors were notified and the activity was stopped. The correct contingency plan was generated and executed or repairs were made to processing equipment. Appropriate condition reports were written so the problems could be evaluated.by managemen The second phase of the SFP clean out project, which was ongoing during this inspection period, included the following activities for each shipment:

1) Filling the inner basket: transporting control road blades (CRBs) to the reactor vessel, crushing them, then loading them into a basket along with a container of Stellite rollers which had been cut off the CRBs before crushing. Also, transporting other items to the reactor to fill the basket (CRB spud ends, incore detectors, and other radioactive reactor hardware).

2) Filling the cask: removing the CRB crusher from the reactor vessel, dose profiling the bottom of the cask liner, placing the shipping cask into the reactor vessel, inserting the liner into the shipping cask, placing the basket into the liner, placing the liner closure lid on the liner, dose profiling the upper portion of the liner, and placing the shipping cask closure lid on the shipping cas ) Draining and deconning the cask: removing the cask from the reactor vessel, draining, surveying and moving the cask from the reactor deck to the equipment lock lay down area, decontaminating, and re-surveying to determine that radioactive material shipping requirements were me ) Placing the cask on the truck: removing the cask from the containment, transporting ;

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to the turbine building truck bay, loading onto the shipping truck, and surveying and smearing the cask again to ensure that the department of transportation regulations were me .

5) After radioactive material shipping requirements were met, the shipping cask was transported to the burial sit i During this inspection period three shipping casks were loaded and transported to the burial site. Each shipping cask held 13 CRBs. To date,39 of 91 CRBs have been shipped to the burial sit c. Conclucion Good worker performance was observed during the second phase of the SFP clean out j project. NRC and Department of Transportation requirements were me .4 Operational Safety Verification (71707)

On June 7,1999, the operations department established the final shift staff complemen The operations shifts are now staffed with one shift supervisor who is a certified fuel j handler, and one decommission!ng worker class 1. No work is being assigned to the i shifts. This is the two person shift authorized by the DTS for non-fuel handling condition I

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. . Gpent Fuel Safety Inspection Scooe (603Q1)

The inspector evaluated spent fuel and fuel pool safety. Factors considered in the evaluation included SFP heatup rate; SFP instrumentation, alarms, and leakage detection; SFP chemistry and cleanliness control; criticality controls; and SFP operation and power supply. The inspector reviewed station logs and held discussions with the licensee. The inspector also inspected the SFP and SFP support system's valve lineups during plant tours, Observations and Findinas

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The inspector reviewed the Auxiliary Operator logs containing SFP parameters and locally monitored SFP level and temperature. The inspector verified the criticality monitor was functioning. The inspector also observed that foreign material controls q were being used in and around the SF The operating range for temperature of the SFP is 40 to 80 *F. The DTS SFP upper ,

temperature limit is 140*F. On June 8,1999, the SFP temperature was 65*F and the i time for the temperature to rise to 140*F was 13.6 days. The present SFP heatup rate is 0.23*F/ hour. During this inspection period the SFP altemate cooling system was shut off during the time when the SFP clean out project activities were in progress to reduce

. the amount of corrosion products potentially getting into the system. Controlling the temperature of the SFP within the operating band with this type of operation was not a proble During this inspection, Field Change FC-699 was implemented to increase the number of plates in the SFP heat exchangers; each heat exchanger originally had 36 plates and that was increased to 76 plates. With an 80 *F service water temperature, the heat exchanger will now maintain a SFP temperature of 95 *F. This will ensure ample cooling capacity during the warm summer months. The previous size heat exchangers met the design specifications for the project but an error was made in the calculations to develop those specifications, resulting in the heat exchangers having about half the number of plates determined as needed when the calculation was corrected. The work was completed on May 5,1999, and the new SFP cooling system was declared operabl Conclusi The safety of the fuel in the SFP was being maintaine .0 Radiological Sa.foty General

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The inspector conducted reviews of ongoing activities in order to assess the overail RP program. Specific findings are detailed in the sections belo .

. Occuoational Radiation Exposure (83750)

Numerous aspects of licensee processes to minimize occupational radiation exposure were selectively examined in order to evaluate overall radiation safety and to provide for early identification of potential problems. Areas examined included: audits and appraisals; planning and preparation; training and qualifications of personnel; external exposure control; intemal exposure controi; control of radioactive materials and contamination; surveys and monitoring; and maintaining occupational exposure ALAR No concerns were identifie .3 Processino and Removal of Materials from the SFP a. Insoection Scooe (83750. 83728)

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The inspector reviewed the licensee's program for planning, processing and removal of materials (CRBs, incore detectors, etc.) from the SFP, as described in Paragraph 2.3, above. During the inspection the inspector reviewed the following radiation work permits '

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1.13 RWP .899002 Operation personnel performing rounds, inspections, tagging and training activitie .14 RWP B99004 Tours, inspections, evaluations and job planning, security functions except coverage of specific RWP .15 RWP B993025 Rigging / loading / handling / inspection and mockup of shipping cas .16 RWP B993029 Repair of underwater equipmen .17 RWP B99033 Processing CABS and incore detectors in SFP/ Reactor Vessel for shipping and disposa The inspector attended the pre-job, RWP, and ALARA briefings for jobs requiring the use of these RWP b. Observations and Findinas The ALARA and the RWP briefings addressed the anticipated radiological conditions, access control, and contamination control and hot particle issues. The licensee also generated specific dose goals for each RWP which was job task evaluated. For ALARA purposes (keeping unnecessary personnel off the reactor deck area during the SFP clean out project) RWP B99004 was rewritten to remove the reactor deck from the general tour RWP. Several different RWPs were written to utilize the different ALARA considerations related to various tasks for the SFP clean out project. This was a good ALARA practice. Briefings improved as experience was gained, and the assignments of workers to the job tasks, and the communication of the work scope to the ALARA planners also improve On June 7,1999, the TV monitor for monitoring the SFP Clean Out Project activities was placed in the computer room out side the controlled area. This is an ALARA item

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which allows management oversite of the project without the managers entering the controlled area, thus saving dose. There are two zoom, pan and tilt cameras which allow viewing activities in the SFP, in the reactor vessel and on the reactor dec Conclusions At the end of this inspection period, ALARA planning, and RWP briefings for job tasks were going well, and remote TV monitoring of SFP activities was provide .4 Radwaste Treatment. and Effluent and Environmental Monitorina (84750)

The inspection included an evaluation of various aspects of licensee activities in the areas of waste treatment and effluent and environmental monitoring to ensure that radioactive waste treatment systerr.., were being maintained and operated to keep onsite and offsite doses ALARA; to ensure that the licensee effectively controlled, I

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monitored, and quantified releases of radioactive materials to the environment; and to ensure that required radiological environmental monitoring programs were effectively implemented. No problems were noted in review of these activitie .5 Solid Radwaste Manaaemont and Transoortation of Radioactive Materials (86750)

4. General ,

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The inspection included an evaluation to determine whether the licensee properly processed, packaged, stored, and shipped radioactive materials, in order to assess the potential for safety problems resulting from these activities and from the transportation !

of radioactive materials. No problems were noted in the evaluation of this are )

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Radiological control activities in the areas of effluents and solid radwaste were being conducted safely and in compliance with applicable requirement ,

i 4.5.2 Packina and Shionina the First Shionina Cask (86740. 86750) Inggpction Scope (86740. 86750)

The inspector evaluated licensee compliance with NRC and Department Of Transportation (DOT) regulations for packaging and shipment of radioactive material Specifically, the inspection focused on the shipment of a significant quantity of licensed material inside a Type B shipping cask. Areas examined included: 1) shipping documentation; 2) loading of the cask within the reactor vessel; 3) closure of the cask for transport; 4) daily RWP briefings; and 5) Radiation Survey Technique : Observations and Findinas Shipping Documentation:  ;

This inspection focused on the first of a planned seven shipments of reactor components for low level radioactive waste disposal. The licensee is using a cask provided by General Atomics to transport the components for disposal. The cask is designated FSV-1 (designed and used in Fort St. Vrain decommissioning) and it is a Type B shipping cask with Certificate of Conformance (CoC) No. USA /9277. The i

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licensee contracted with NUKEM to manage the loading of the cask and to assist the licensee's waste management group for the shipment of the waste. This particular shipment included:

13 Control Rod Blades "

2 full Incore detectors 52 Stellite rollers and 5 spud ends in order to determine the nuclides and activity of the above components, the licensee contracted with Waste Management Group (WMG) to characterize the above components. Information provided to WMG includes, but is not limited to, dose profiles of the above components, metallurgic analysis of the components and estimated time and conditions while the components were being irradiated within the reactor vesse The characterization is provided to the licensee to be used in the completion of waste documents for the disposal of the waste which includes, but is not limited to, nuclides present, activities of each nuclide and characterization of the waste class. According to ;

figures provided by WMG, the above components possessed a total of approximately i 7.94E+06 millicuries (2.94E+08 megabecequeris). The inspector reviewed the licensee's interaction with WMG, reviewed the procedures for the classification of waste i and monitored the interaction betweeri the licensee and Bamwell Waste Management Fac!!ty, operated by Chem-Nuclear Systems, and the State of South Carolina. The licensee appeared to have adequately monitored the interaction between the above contractors to onsure compliance with NRC and DOT regulation CFR 172.201(d), referenced by 10 CFR 71.5, requires, in part, that shipping papers must contain an emergency response' telephone number. The NRC expects (Inspection Procedure 86750) that, as general guidance, emergency responders will be able to provide emergency response information within 15 minute The number for the phone at the plant monitoring station (formerly the plant control room) was listed as the emergency number on shipping papers for this shipment of radioactive material. The inspector interviewed staff manning the monitoring station !

(which is occupied 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day,7 days a week) and found that they had difficulty obtaining the appropriate emergency response information in a mock scenario presented by the inspector. This suggested these individuals may not have been able to provide the required information to an emergency responder within 15 minute I According to the licensee, all monitoring station operators were trained on thek )

responcibilities reganiing emergency response information earlier in the year. Licensee staff in'dicated that they would look into the matter to determine whether further training

. is necessary or other solutions should be considered for future shipments to ensure operators can promptly provide the appropriate informatio During the onsite inspection, the inspector was informed that the licensee identified that the CoC Number, as documented on the cask, was the incorrect CoC Numbe '

Specifically, labeling on the cask stated the CoC Number was USA /6346 when the CoC Number was actually USA /9277. The licensee confirmer .vith the NRC and General Atomics that the CoC Number was indeed USA /9277 a ..; received guidance from General Atomics on how to correct the problem in a letter dated April 30,1999. The licensee performed the required modification prior to the shipment of the cask on May 7, 199 .

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Loading of the Cask Within the Reactor Vessel:

During the inspection, the inspector observed several parts of the implementation of Procedure No. NUKEM 1874-010-1036, Revision 2, " HANDLING AND LOADING THE FSV-1 SHIPPING CASK." Specifically, the inspector observed or reviewed the portion of the procedures surrounding the loading of the liners and irradiated material inside the liners. No problems were note Closure the Cask for Transport:

Closure of the cask for trani port is also govemed by Procedure No. NUKEM 1874-010-1036. The procedure includes instructions conceming the torquing of 24 bolts on the head of a cask and then pressure (leak) testing the cask for a secure cask seal. The torque specification was 950 - 1000 lb-f The inspector noted that the licensee was utilizing a 100 - 600 lb-ft torque wrench to an indicated 250 lb-ft of torque. The wrench was being used in combination with a mechanically geared " torque multiplier" to increase applied torque by a ratio of 4 to 1 to achieve the specified torque requirements. According to the licensee, the torque wrench had been calibrated within the preceding 12 months but the torque multiplier had not been calibrated to verify that the multiplication factor was 4 to 1. The inspector requested that the torque multiplier value be verified prior to the shipment of the cas The licensee subsequently reported to the inspector that a series of post-Job checks showed the torque multiplication factor was not 4 to 1 as previously believed, but was consistently in the range of 3.4 to 3.5 to 1. Examination of the vendor data on the device established that the gear ratio was a 4 to 1 ratio, but that the specified torque multiplication factor was 3.4 to 1. Over time, the licensee had lost track of the actual performance capability of the device. In addition, the licensee's review determined that a torque multiplier may have an inherent mechanical errer range of up to 15 percent and the multiplication factor may not be linear over a wide range of torque values. This placed some doubt on the propriety of using such a tool for a job with a 5 percent accuracy specification without a post-Job accuracy verificatio As a result of these fhdings, the cask bolts needed to be re-torqued to meet cask closure specifications. This was done, and both pre-job and post-job verifications demonstrated torque accuracy. According to the licensee, the initial under-torque of the bolts did not damage the gasket on the cask lid and a pressure test of the cask after re-torquing did not identify a seal failure. The licen.wa instituted corrective actions to ensure the identified problem will not recu Daily Radiation Work Permit Briefings:

Prior to each day of work, the licensee holds a RWP briefing to discuss items which

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occurred the previous day, current status of the project and what the goals are for the day. The inspector attended each briefing while on site ano noted that the licensee's staff adequately characterized the daily duties and responsibilities of various staff members, addressed unexpected work items or delays and addressed concems from staff members regarding procedures and radiation safet l

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Radiation Survey Techniques:

The inspector observed radiation direct measurement and contamination sur/ey techniques performed by the RP staf The licensee was using an appropriate survey meter to measure the radiation levels on the cask as it exited containment on May 5,1999. The inspector also observed that the RP technician was performing radiation measurements in locations where the highest j fields were expected and was performing such measurements in accordance with standard radiation survey techniques. The inspector took comparison radiation measurements and found measured radiation levels were comparable; within plus or minus 20 percent of the values found by the license The inspector also reviewed the contamination survey techniques which included proper documentation and control of wipes, the wearing of proper protective clothing and standard radiation safety practices. The contamination survey of the cask which the inspector observed was performed after the cask had been removed from the containment building and placed in the turbine building for the evening. The inspector independently collected wipe samples from the cask which were counted by the NRC Region lli laboratory to determine if significant differences were noted between the wipe results of the licensee and NRC. No significant differences were noted between the licensee's and NRC's resuhs; allwipe samples showed removable contamination levels well below the specified DOT limit of 2200 dpm/100 square centimeters, and below the licensee's administrative limit of 1000 dpm/100 square centimeter The inspector reviewed licensee documentation conceming the first FSV-1 cask shipment, on May 7,1999, and for a Low Specific Activity / Surface Contaminated Object shipment which occurred on May 6,1999.' The review considered contamination su:veys of the cask, extemal radiation measurements of the cask, waste manifest form, labeling of the shipment and placarding of the vehicle. The inspector did not identify any issues which warranted further NRC revie As part of the licensee's radiological clearance of material from the site, Procedure RM-56 states, in part, that vehicles leaving the plant perimeter fence should be surveyed based upon the vehicle's use and location onsite. On May 5,1999, the ,

inspector observed an RP technician performing a survey of a garbage truck which had i entered the plant perimeter fenced area to pick up noreradioactive trasn. The RP )

technician's practices for performing this radiological sutvay were good. After the survey, the inspector requested that the RP technician demonstrate that the survey meter could detect radiation. A radioactive source used for daily operational checks of the device was used, but survey meter failed to respond.' Evaluation showed that the cable connecting the electronics to the probe was defective. The RP technician ;

immediately took the survey meter out of service and documented that the survey meter 1 was not operational. The inspector reviewed the daily survey meter source check logs and noted that the survey meter had been responding to radiation earlier in the da Licensee management was informed of the item during an pre-brief exit held on May 6, 1999, and agreed with the inspector's assessment that technicians must always use an operable survey instrument when conducting radiological survey ,

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' Conclusion No violations of NRC requirements were identified. However, several items were noteri which merit licensee attention, including: 1) adequacy of knowledge of monitoring station operators regarding emergency response information to be provided if an incident should occur during transport of radioactive material; 2) use of a defective survey instrument while performing a radiological survey; and,3) erroneous use of a torque multiplier, which resulted in the under-torque of the FSV-1 cask bolts. The licensee planned appropriate actions for each of the items abov )

4.5.3 Packina and Shiooina the Second Shiocina Cask (86740. 86750)

l Insoection Scooe (86740. 86750) -

An inspection and evaluation were also made of the loading of the second shipping cask and the documents associated with the packaging and chipe.vnt of the cas Observations and Findinas The icspector observed the loading of the FSV shipping cask onto a transport carrier for shipment to Bamwell in South Carolina. This was the second such shipment from Big Rock Point, and the cask contained over 8000 curies in irradiated components that had been removed from the SFP The radiological controls that were in place during the loading of the cask were observed to be adequate for the activities taking plac Independent surveys of the cask taken by the inspector were compared to the licensee's survey results of the cask. The results compared very favorably. When an individual inadvertently passed a meter across a contamination boundary to someone on the clean side, the Radiation Protection Technician was quick to point out the error and immediately surveyed the meter to ensure no contamination was spread across the

~ boundar The inspector reviewed Procedure NUKEM 1874-010-1036, Handling & Loading the ;

FSV-1 Shipping Cask, Revision 2, along with the Waste Manifest and Bill of Lading I associated with the May 21,1999 FSV cask shipment from Big Rock Point. The procedure task / data sheets and documentation were observed to be properly filled out l

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and adequate to ensure the radioactive waste was being properly characterized and packaged prior to shipment. No problems were noted with any of the records of calculations or of the surveys performe ' Conclusions i

The loading of a FSV shipping cask onto a transport carrier and subsequent shipment to Bamwell in South Carolina was observed to be carried out as required, and with associated documentation and record keeping completed as required by procedure. No problems were identified with the activities surrounding the shipmen .

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I Exit Meeting Summary '

The inspector presented the inspection results to members of licensee management at the conclusion of the inspection on June 8,1999. Tho licensee acknowledged the findings presented. The inspectors reviewed no proprietary information during this inspection perio The licensee did identify any documents or processes reviewed by the inspectors as proprietar PARTIA'. 9 lST OF PERSONS CONTACTED Licensee M. Bourassa, Licensing Supervisor L. Darrah, Technical Support & Assessment Superintendent (RP&ES)

C. Jurgens, Planning, Maintenance & Construction Manager M. Lesinski, Radiation Protection and Environmental Services Manager (RP&ES) l R. McCaleb, Nuclear Performance Assessment, Site Lead (NPAD)

G. Petitjean, Senior Engineer L. Potter, Maintenance Suparvisor K. Powers, Site General Mar,ager J. Rang, Industry issues G. Rowell, Corrective Action Supervisor l

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G. Szczotka, Manager, Nuclear Performance Assessmr ! Department (NPAD)

W. Trubilowicz, Operations, Cost Engineering & Schec dng Manager M. VanAlst, Security Manager R. Wills, Radwaste Superintendent G. Withrow, Engineering & Licensing Manager K. Wooster, Emergency Planning Coordinator E. Zienert, Human Resources & Administrative Department Manager

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INSPECTION PROCEDURES USED i

IP 36801: Organization, Management, ariu Cost Controls at Permanently Shut Down Reactors IP 37801: Safety Review, Design Changes, and Modifications at Permanently Shut Down Reactors '-

IP 40801: Self-Assessment, Auditing, Corrective Action IP 42700: Review of Modifications and Procedures -

IP 60801: Spent Fuel Pool Safety at Permanently Shut Down Reactors l IP 62801: Maintenance and Surveillance at Permanently Shut Down Reactors IP 71707: Operational Safety Verification IP 71801: Decommissioning Performance and Status Review at Permanently Shut Down Reactors -

IP 83750: Occupational Radiation Exposure IP 84750: Radwaste Treatment, Effluent and Environmental Monitoring IP86740: Inspection of Transportation Activities IP 86750: Solid Radwaste Managtment and Transportation of Radioactive Materials l

ITEMS OPENED, CLOSED, AND DISCUSSED  !

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Opened l None

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VIO 50-98004-02 Improper use of and inadequate Reliability Criteria (Maintenance Rule). 4 IFl 50-98002-05 Hazardous waste shipped to waste processing facilit Discussed None-s i

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LIST OF ACRONYMS USED ALARA As-Low-As-Reasonably-Achievable BRP .. Big Rock Point CARB Corrective Action Review Board CoC Certificate of Compliance CFR Code of Federal Regulations

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CRB Control Rod Blade DOT Department of Transportation DE Defueled Emergency Plan DTS Defueled Technical Specifications FHSR' Final Hazards Summary Report l FSV Fort Saint Vrain (Shipping Cask)  !

IFl Inspector Followup item ISSSF Important to Safe Storage of Spent Fuel MRB Management Review Board NPAD Nuclear Performance Assessment Department NRC .

Nuclear Regulatory Commission PSDAR Post-Shutdown Decommissioning Activities Report QA Quality Assurance RP Radiation Protection RWP Rac'iation Work Permit SFP Spent Fuel Pool SRC Safety Review Committee SSCs Systems, Structures and Components TS Technical Specification URSQ Unreviewed Safety Question VIO - Violation WMG Waste Management Group ,

LICENSEE DOCUMENTS REVIEWED NUKEM 1874-010-1036, Handling & Loading the FSV-1 Shipping Cask, Revision Additional licensee documents reviewed and utilized during the course of this inspection are specifically identified in the " Report Details" abov I 20