IR 05000344/1998004

Insp Repts 50-344/98-04 & 72-0017/98-01 on 981106-07 & 1202-10.Violations Noted.Major Areas Inspected: Decommissioning & Dismantlement Activities at Plant Site

ML20202C223
Person / Time
Site:	Trojan  File:Portland General Electric icon.png
Issue date:	01/21/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20202C167	List: IR 05000344/1998004 ML20202C162 ML20202C211
References
50-344-98-04, 50-344-98-4, 72-0017-98-01, 72-17-98-1, NUDOCS 9901290337
Download: ML20202C223 (20)

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ENCLOSURE 2 U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket No.: 50-344;72-17 License No.: NPF-1 Report No.: 50-344/98-04; 72-17/98-01 Licensee: Portland General Electric Company Facility: Trojan Nuclear Plant Location: 121 S. W. Salmon Street, TB-17 Portland, Oregon Dates: November 6-7,1998 and December 2-10,1998 Inspector: J. V. Everett, Sr. Health Physics inspector Division of Nuclear Materials Safety Fuel Cycle & Decommissioning Branch M. P. Shannon, Sr. Radiation Specialist Division of Reactor Safety Plant Support Branch L. H. Thonus, Project Manager Office of Nuclear Reactor Regulation Non-Power Reactors and Decommissioning Project Directorate H. K. Lathrop, Transportation and Storage Safety inspector Office of Nuclear Material Safety and Safeguards Spent Fuel Projects Office Approved By: D. Blair Spitzberg, Ph.D., Chief Fuel Cycle & Decommissioning Branch Attachment: Supplemental Information

~9901290337 990121 PDR ADOCK 05000344 G PDR

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-2-EXECUTIVE SUMMARY Trojan Nuclear Plant NRC Inspection Report 50-344/98-04; 72-17/98-01

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Decommissioning and dismantlement activities at the Trojan site were aggressively moving forward. Recent approvals from the NRC, Department of Transportation, and State of Washington to barge the intact reactor vessel to Hanford, Washington for burial was a major step forward in the decommissioning effort. Plans to bury the reactor vessel intact will reduce the overall public and worker exposure to this major dismantlement activity compared to the alternative of segmenting the reactor. In preparation for shipment, the reactor vessel was filled with low density concrete which will fix any loss contamination inside the vessel. The reactor vessel is scheduled to be lifted out of containment in June 1999 and shipped to Hanford in July 199 Dry cask storage activities were also progressing well with plans to begin fuel movement in April 1999. To support this date a number of licensing issues remain to be completed. Two major site activities include a January 1999 dry run demonstrating the capability of the licensee to respond to a leaking cask and to transfer the cask to an overpack. The second major dry run will demonstrate the licensee's capability to load a cask in the spent fuel pool and transfer the cask to the Independent Spent Fuel Storage Installation (ISFSI) pad. This will be conducted in March 1999. Successful completion of these two dry runs and completion of the remaining licensing issues could allow fuelloading of the casks at Trojan as early as April 199 Decommissionino Performance and Status Review

The Trojan reactor vessel was successfully filled with low density cellular concrete in preparations for shipment and burial at Hanford, Washington. Work activities were completed with no significant problems. Concrete was observed coming out of the reactor vent head indicating that the vessel had completely filled with concrete (Section 1).

Decommissioning work activities, fire loading, safe work practices and radiological controls were observed during plant tours. No unsafe conditions were noted (Section 1).

Oraanization. Manaaement. and Cost Controls

Several revisions to the Trojan Permanently Defueled Emergency Plan and procedures had been completed in 1998. The changes were mostly editorial or provided clarification. The changes did not decrease the effectiveness of the emergency response program (Section 2).

Spent Fuel Pool Safety

Spent fuel pool water level, temperature, and boron concentration were being maintained within technical specifications. The new modular cooling system was providing adequate cooling to keep the spent fuel pool water temperatures below the

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technical specification limit. Operations procedures had been revised to reflect the new l cooling and cleanup system for the spent fuel pool (Section 3). l

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! Occupational Radiation Exposure

• A good external radiation control program was maintained by the licensee. Station workers used the personnel contamination monitoring equipment properly. High ,

radiation areas were properly controlled and posted. Radiation workers properly wore

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personal dosimetry (Section 4).

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• Whole-body counters were calibrated using standards traceable to the National Institute *

I of Standards and Technology. Air sampling equipment was appropriately placed to monitor airborne conditions in the work areas. High efficiency particulate air filter j ventilation units were used to limit airborne exposures to workers. Personnel i

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contamination events were properly responded to and recorded in accordance with procedural requirements (section 4).  !

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! * Radioactive material containers were labeled, posted, and controlled. All laundry and trash containers were properly maintained. Contaminated areas were posted and clearly identified. Portable radiation protection instrumentation observed in use in the l radiological controlled area had been calibrated and source response checked. One  !'

j violation was identified for the failure to maintain posting of an airborne radioactivity area l as required by 10CFR20. Based on the corrective actions identified and taken by the  ;

L licensee, no response to this violation is required (Section 4).  ;

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• The ALARA planning program was effectively implemented. Radiation protection j ALARA personnel were appropriately involved with the planning and monitoring of the l reactor vessel removal project (Section 4).  ;

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! Radwaste. Effluents. & Environmental Proarams

-* Walkdowns of the gaseous and liquid effluent monitors and radioactive waste storage i tanks were conducted. All radioactive effluent monitors were operational, calibrated and I properly maintained. Waste storage tanks and associated pumps were in good physical

! condition with no visible leakage noted (Section 5).  ;

independent Spent Fuel Storace installation

• The licensee completed the construction of the first TranStor concrete cask in accordance with approved procedures. The concrete met the requirements specified for the cask. No discrepancies or concerns were identified with the construction activities (Section 6).

• Plans for loading spent fuel into dry cask storage were aggressively moving forwar Two major walk-throughs of cask loading and unloading procedures were planned for January and March,1999 with possible fuel loading as early as April 1999 (Section 6).

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-4-Report Details Summarv of Plant Status Dismantlement and decommissioning activities at Trojan continued to progress at a rapid pac Staffing levels were high, with over 400 personnel onsite. Staffing levels reflected four major efforts underway. These included the reactor vessel removal project, continued decommissioning of the site, activities related to dry cask storage of the spent fuel, and preparations for the final radiological survey of the sit The reactor vessel removal project took a major step forward with the approval by the NRC, U. S. Department of Transportation and State of Washington to allow transport and burial of the reactor vessel at Hanford, Washington. In preparations for shipment of the reactor vessel, the licensee filled the vessel with low density concrete to fix any loose contamination inside. This involved two concrete pours approximately one week apart. Both efforts were completed with no significant problems. NRC inspectors observed both concrete pours and visually confirmed during the second pour that concrete was overflowing out of the reactor vent head on top of the reactor vessel as planned to ensure the process was complet Work activities at the ISFSI were progressing toward a projected fuel loading in the spring of 1999. The onsite construction of the first concrete storage cask was observed by the NR Work activities and the licensee's quality control oversight of activities were found to be acceptable.. The concrete cask, a basket, a transfer cask, and an overpack were located at the l ISFSI and being used by the licensee to practice for fuel loading. The transfer station was erected on the ISFSI pad and was being used during the practice activities. The security fence had been installed. The remaining security systems were being c'ompleted and tested.

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The Oregon Office of Energy conducted a public meeting on December 9,'1998 in Portland, Oregon to solicit comment on draf t rules being developed by the State of Oregon concerning the approval to allow dry cask spent fuel storage at the Trojan site. Two members of the public attended the meeting. The meeting provided an opportunity for the licensee and the Oregon Office of Energy to discuss the plans for dry cask storage at Trojan with the publi Decommissioning Performance and Status Review (71801)

j Reactor Vessel Removal Project Inspection Scope On October 22,1998, November 23,1998, and November 24,1998, PGE received permission from the NRC, U. S. Department of Transportation, and State of

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Washington, respectively, to transport and bury the Trojan reactor vessel at Hanford,

! Washington. A major step in the process to prepare the vessel for shipment and burial involved the filling of the reactor vessel witn low density cellular concrete. The work

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activities related to this project were observed to verify that the reactor vessel had been successfully filled.

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! Observation and Findinas l

On December 3,1998, the licensee poured the first of two lifts of low density cellular t concrete into the reactor vessel. Pre-job walk-throughs were conducted in the

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containment and at the concrete batch plant. The pre-job briefing and safety meeting

[ was attended. All preparations were found to be appropriate for the work planne ;

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l Concrete work activities were implemented according to PGE maintenance request, MR 17690, Revision 1. Contingency plans were developed should problems occur during the concrete pouring activities. The maintenance request and contingency plans were reviewed and found acceptable. The concrete batching, density measurements, t

. and pouring of the first lift were observed by an NRC inspector. No concerns were

! identified during the work effor The pre-job walk through in containment and at the concrete batch plant were thorough and attended by both the persons to be performing the work and the quality control personnel who would perform independent checks and verifications. The equipment to ,

be used was pre-staged and labeled. Hoses were walked down and independently '

i verified by quality control personnel. The combined pre-job briefing and safety meeting i

covered the job scope, planned sequence of activities, communications systems, j designation cf individuals in charge of each segment of activity, contingency plans and i safety requirements for each work zone.

On December 3,1998, the first lift was completed in tecordance with MR 17690. The

reactor vessel was filled to the level of the inlet nozzles. All concrete batches met the

! criterion of between 45 and 65 lb/ft specified in NRC's approval document for the I reactor vessel package. No concrete batches were outside the 50-60 lb/ft range. Each

! batch was tested prior to being discharged to the hopper which fed the screw pump

! which in turn pumped the concrete into the reactor vessel. The lif t was well coordinated and proceeded uninterrupted until completion at the desired elevation of the reactor vessel nozzles as evidenced by flow out an indicator valve and clear section of tubing designed for that purpose. The licensee had developed contingency plans for batches

found out of specification, piping leakage, flow blockage, and malfunctions of the l concrete batch plant. None of these contingency plans were needed during the work effort.

l The second concrete pour occurred on December 10,1998. Concrete was injected in

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an inlet and outlet nozzle and a control rod drive mechanism penetration. The reactor vessel was filled until concrete was observed in the tubing attached to the reactor vent head. Approximately 5300 ft 2of concrete was injected into the reactor vessel. The reactor vessel was to vent for 28 days to allow the excess moisture from the concrete to escape and the reactor vessel and concrete to reach equilibrium condition Conclusion

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The Trojan reactor vessel was successfully filled with low density cellular concrete in !

preparations for shipment and burial at Hanford, Washington. Work activities were  ;

completed with no significant problems. Concrete was observed coming out of the {

reactor vent head indicating that the vessel had completely filled with concret l

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6-1.2 Status of Decommissionina. Plant Tours and Conduct of Activities a. Inspection Scope The licensee was actively completing dismantlement of a number of systems. A plant tour was conducted to evaluate housekeeping, fire loading, safe work practices and radiological controls, b. Observations and Findinas Decommissioning work continues for the containment, fuel building, and auxiliary building. Staffing levels were over 400 personnel, reflecting a very active level of wor Numerous major work activities were underway with a significant amount of work planned for the next several months. Several new areas were observed to be under dismantlement since the last inspection. Equipment in the containment had been significantly reduced. There was little remaining other than the reactor vessel and concrete. The systems connected to the spent fuel pool had been disconnected and the ,

pool was operating on the modular spent fuel pool cooling and clean-up systems. The '

facility status was consister,t with the Defueled Safety Analysis Report, Revision 7, and !

the Decommissioning Plan, Revision '

A tour of the control room was conducted. The shift manager was on duty which met j the requirement in Technical Specification 5.2.2.(a) and (b) for control room staffin l The shift manager provided a tour of the active panels and readouts in the control roo l This included the area radiation monitors for the spent fuel pool and the ventilation system monitors. The control room logs were reviewed. Daily surveillances were being conducted of the radiation protection system instrumentation in the control roo A plant tour was conducted to observe work activities underway and the condition of the I facility. Areas were found to be properly posted and barricaded. Radioactive material that had been collected during work activities was properly bagged, labeled and placed to the side out of the way. No fire loading or safety issues were observed during the tour Conclusion l Decommissioning work activities, fire loading, safe work practices and radiological controls were observed during plant tours. No unsafe conditions were note Organization, Management, and Cost Controls (36801)

2.1 Inspection Scoce Several revisions to the permanently defueled emergency plan and procedures were issued during 1998. A review was completed of the revisions to verify that the emergency program was not reduced in effectiveness. Arrangements for medical care for potential severe head trauma injuries were also discussed with the license . . - .. .

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-7-2.2 Observation and Findina Revisions 6 and 7 to the emergency plan were issued July 1998 and December 1998, respectively. Revision 6 removed the letters of agreement from Appendix A of the emergency plan and replaced the letters with a list of the organizations in which agreement letters were maintained. The current letters of agreement were being maintained in a file by the emergency preparedness supervisor. The letters of agreement had been reviewed by the licensee in June,1998. The licensee had contacted each of the agencies requesting cenfirmation that the information in the letters of agreement was current. Other changes made as part of Revision 6 were editorial. Revision 7 to the emergency plan consisted of an update to Section 8. describing the process and effluent radiation monitors. The revision reflected the i current installed condition of the monitors resulting from changes made during decommissionin Emergency Plan Implementing Procedure (EPIP) No.1 " Classification of Emergencies,"

Revision 5 was issued August 13,1998. Several definitions were revised and clarifie Emergency Plan implementing Procedure (EPIP) No. 2 * Notifications," Revision 5 was issued August 13,1998. This revision included changing the name of the Oregon Department of Energy to the Office of Energy and providing a note on an alternate notification method if the primary notification method fail Emergency Plan Implementing Procedure (EPlP) No. 3 " Response Organization Checklist," Revision 7 was issued September 17,1998. The Oregon Office of Energy title was revised. A note was added to the maintenance coordinators checklist to assign i two or more persons to emergency teams entering unknown or possible hazardous j situation Emergency Plan Implementing Procedure (EPIP) No. 5 " Emergency Preparedness Test Program," Revision 9 was issued September 17,1998. The changes were minor editorial and clarification change None of the changes to the emergency preparedness program were identified as having reduced the effectiveness to the emergency plan, procedures, or response capabilit The Trojan Defueled Emergency Plan, Section 5.5 described the arrangements for medical support to an injured and contaminated individual. St. John Medical Center in i

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Longview, Washington, was the agreement hospital for receiving injured personnel from Trojan who may be contaminated. This hospital was the only local hospital available ,

and has a long history of working with Trojan, including years of participation in !

emergency medical drills. However, St. John Medical Center was not staffed or equipped to handle severe head trauma cases. This limitation was discussed with the

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licensee's emergency preparedness coordinator to determine the contingency plans l established for this type of injury. Three hospitals had been identified by the licensee as alternative hospitals for severe head trauma injury. These were Oregon Health Science l

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Medical Center in Vancouver, Washington. All three hospitals had confirmed with the j i

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licensee's emergency preparedness coordinator that they were prepared to accept I contaminated and injured persons as part of their hazmat emergency program, including .

severe head trauma injuries. The licensee had offered to provide specialized training to !

the hospitals. The hospitals were considering the licensee's offe j l Conclusion ,

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Several revisions to the Trojan Permanently Defueled Emergency Plan and procedures had been completed in 1998. The changes were mostly editorial or provided  ;

clarification. The changes did not decrease the effectiveness of the emergency response progra Spent Fuel Pool Safety (60801,86700) Inspection Scope ,

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Compliance with requirements and commitments for the spent fuel pool in the Trojan !

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Technical Specifications, "Defueled Safety Analysis Report," and the licensee's response to NRC Bulletin 94-01 " Potential Fuel Draindown Caused by inadequate ;

Maintenance Practices at Dresden Unit 1," were reviewed. Operation of the new i modular spent fuel pool cooling system was evaluated to confirm the system was !

operating with sufficient margin for water temperature such that the technical specifications for maximum water temperature in the spent fuel pool would not be

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i Technical Specification 3.1.1 required the spent fuel pool water level to be maintained )

223 feet over the top of the irradiated fuel assemblies. The computerized trending j records for the period of August 1 through December 7,1998, were reviewed. Water ;

level was typically maintained between 24 feet and 24 feet 6 inches. At no time was l

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- water level below 23 feet 6 inche Technical Specification 3.1.2 required the spent fuel pool boron level to be maintained 22000 parts /million (ppm). Surveillance record, form C-258, for the spent fuel pool was reviewed for the period of September 28 through December 7,1998. Weekly surveillances were conducted. No boron samples were measured below the required l 2000 ppm limi Technical Specification 3.1.3 required the spent fuel pool temperature to be maintained s140*F. The computerized trending records for the period of August 1 through December 7,1998, were reviewed. The highest temperature recorded was 122*F on

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August 16,1998. By September, temperatures were consistently below 100*F. For November and December, typically temperatures were around 85*F.

l The new modular cooling system for the spent fuel pool began operation July 199 From January 1 through June 30,1998, spent fuel pool temperatures remained below

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-9-75'F. After the system began operation, the spent fuel pool temperatures increased due to the passive cooling process used by the new system. Outside temperatures effect the pool temperature because outside air was used for cooling the coils. Despite hot temperatures in July and August, the new system was able to maintain cooling of the spent fuel pool below the required technical specification limi With the new modular cooling system installed and operational, the systems that had previously provided cooling and cleanup of the spent fuel pool were dismantle Procedures ONI 4-4, "Off-Normal Instructions - Spent Fuel Pool System Trouble,"

Revision 11 dated September 14,1998; 014-3 " Service Water System Operating instructions," Revision 40 dated September 23,1998; and FHP-18 * Fuel Movement and Position Verification," Revision 5 dated December 1,1998, were reviewed to verify that changes to the spent fuel pool systems had been incorporated into procedures. The reviewed procedures were found to be current with the changes made to the spent fuel pool system The licensee's June 30,1994 response to NRC Bulletin 94-01, " Potential Fuel Draindown Caused by inadequate Maintenance Practices at Dresden Unit 1," was reviewed to verify that changes to the system were consistent with commitments made in the licensee's 1994 letter to the NRC. Statements concerning water level alarms, procedures for response to low water problems, and procedures for adding make-up water to the spent fuel pool were found to be current. Availability of systems to provide make-up water to the spent fuel pool was confirme The Trojan Defueled Safety Analysis Report, Revision 7, included a requirement in Section 3.2.2, " Spent Fuel Pool and Fuel Storage Racks," to administratively restrict storage of fuel assemblies in the cells of the remaining racks immediately adjacent to the opening created by the removal of one of the racks. Procedure FHP-18," Fuel

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Movement and Position Verification," Revision 5 dated December 1,1998, was reviewed to confirm the requirement to restrict storage of fuel at that location. Section 4.9 of the procedure specifically restricts storage of fuel assemblies adjacent to or in the opening created by the removal of Rack No. .3 Conclusion Spent fuel pool water level, temperature, and boron concentration were being maintained within technical specifications. The new modular cooling system was providing adequate cooling to keep the spent fuel pool water temperatures below the technical specification limit. Operations procedures had been revised to reflect the new cooling and cleanup system for the spent fuel poo _-__-___

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- 10-4 Occupational Radiation Exposure (83750) External Exoosure Controls Inspection Scope Selected radiation workers and radiation protection personnel involved in the external exposure control program were interviewed. A number of tours of the radiological controlled area were performed. Areas reviewed included radiological controlled area access controls, control of high radiation areas, and dosimetry us Observations and Findinos Activities at the radiological controlled area access / egress control point were observe Station workers were using the personnel contamination monitoring equipment properl Radiation protection personnel provided timely response and direction to station workers who alarmed the personnel contamination monitor .

During tours of the radiological controlled area, high radiation areas were observed to j be properly controlled and posted. Based on interviews with radiation protection j personnel and review of radiological area survey maps the inspectors noted that the '

licensee had eliminated all radiological components and/or equipment that required areas to be controlled as locked high radiation areas. Locked high radiation areas are areas with greater than 1000 millirem /hr radiation at 30 centimeter All personnel in the radiologically controlled area were observed wearing their personal dosimetry properly. Workers questioned knew to contact radiation protection personnel if their electronic dosimeters alarme !

c. Conclusions A good external radiation control program was maintained by the licensee. Station workers used the personnel contamination monitoring equipment properly. High radiation areas were properly controlled and posted. Radiation workers properly wore personal dosimetr .2 Internal Exposure Controls Inspection Scope Selected radiation protection personnel involved with the internal exposure control program were interviewed. Areas discussed included the whole-body counter calibration program, air sampling program, including the use of continuous air monitors and filtration Units, and personnel contamination progra l

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-11 - Observations and Findinas The whole body counting equipment was calibrated semi-annually by a vendo Whole-body counters were verified to be calibrated using standards traceable to the NationalInstitute of Standards and Technology. An acceptable phantom was being used by the licensee along with radiation sources that covered energy ranges between approximately 356 -1836 kev. The inspectors concluded that a proper whole-body cc nmg calibration program was in plac ,

The licensee used high efficiency particulate air filter ventilation units to reduce airborne exposures to workers. Local job specific air sampling was used for specific jobs to monitor airborne radioactivity levels. During tours of the radiological controlled area, job coverage air sampling was observed to be properly placed to monitor airborne conditions at the work area. Additionally, continuous air monitors (CAMS) were properly used throughout the radiological controlled area. However, the inspectors observed that the flow gauge associated with the CAMS was by-passed and not indicating the amount of air flow across the detector surface. The licensee was questioned on how they determined if the CAMS were operating properly without the flow indicator working. The radiation protection supervision explained that the CAMS were used as an indication of an airborne condition and not to determine the airborne concentration in an are Though this may be the intended use of the CAMS, with the flow gauge by-passed, there was no indication that the CAM was operating properly. The licensee agreed and placed the CAM flow gauges back in service within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Because of the proper placement of job specific air samples, no actual safety consequences were associated with the above observatio From observations made at the egress point of the radiological controlled area and interviews of radiation protection personnel, the inspectors determined that personnel contamination events were properly handled and recorded in accordance with procedural requirements. Several personnel contamination event records were selected for review. These events involved radiation workers with low level external contamination in the facial area. Whole-body counts had been conducted as required by station procedures. No problems were identified with the personnel contamination progra c. Conclusions Whole-body counters were calibrated using standards traceable to the National Institute ,

of Standards and Technology. Air sampling equipment was appropriately placed to l monitor airborne conditions in the work areas. High efficiency particulate air filter j ventilation units were used to limit airborne exposures to workers. Personnel contamination events were properly responded to and recorded in accordance with procedural requirements.

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-12- Control of Radioactiva Materials and Contamination: Survevina and Monitorina Inspection Scope Control of radioactive material, portable instrurnentation calibration and performance checking programs, adequacy of the surveys necessary to assess personnel exposure, and posting of airborne radioactivity areas were reviewe Observations and Findinas During tours of the radiological controlled area, all radioactive material containers were observed to be labeled, posted, and controlled. All laundry and trash containers were properly maintained. Contaminated areas were posted and clearly identifie Instrumentation observed in use in the radiological controlled area was calibrated and source response checked in accordance with station procedure Independent radiological measurements performed by the inspectors confirmed that area radiological postings reflected general radiological conditions within the facility, in general, all radiological postings were conspicuous and clearly posted. However, on December 9,1998, during the review of airborne radioactivity sampling logs, the inspectors noted that on December 8,1998, at approximately 11:15 a.m., the licensee identified that an air sample taken under the reactor vessel, in the containment building, was 1.2 derived air concentrations (DAC). In accordance with licensee Procedure RP-143, * Posting of Radiologically Controlled Areas," and 10 CFR 20.1902(d), the area required posting as an airborne radioactivity area. The area was properly posted as an airborne radioactivity area at the time sampling was conducted. At approximately 4:10 p.m. the same day a second air sample was taken. The results of the second sample revealed 1.8 DAC, which required continued posting of the area as an airborne radioactivity area. On December 9,1998, the licensee removed the airborne radioactivity area posting, in the area under the reactor vessel, for approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to allow workers to enter the area, without verifying that airborne radiological conditions in the area had dropped below required posting limits. A review of the personnel contamination event log was completed for the workers who entered the area under the reactor vessel during the time the area was not posted. None of the individuals had alarmed the personnel contamination monitor The failure to post the area under the reactor vessel for approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> on December 9,1998, as an airborne radioactivity area was determined to be a violation of 10 CFR 20. This violation is identified as Violation 50-344/9804-01. Specifically, the following requirements of 10 CFR 20 would apply to this situatio Each licensee is required by 10 CFR 20.1501 to conduct surveys necessary for the licensee to comply with the regulations in Part 20 and that are reasonable under the circumstances to evaluate the extent of radiation levels, concentrations or quantities of radioactive materials, and the potential radiological hazards that could be presen ____

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-13-Surveys are defined in 10 CFR 20.1003 as an evaluation of the radiological conditions

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and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive material or other sources of radiatio An airborne radioactivity area is defined in 10 CFR 20.1003 as a room, enclosure, or area in which airborne radioactive materials, composed wholly or partially of licensed material exists, in concentrations in excess of the DACs specified in Appendix B, to S 20.1001-20.2401.10 CFR 20.1902(d) requires the licensee to post each airborne radioactivity area with a conspicuous sign or signs bearing the radiation symbol and the words Caution, Airborne Radioactivity Area or Danger, Airborne Radioactivity Area."

In addition to the requirements of 10 CFR 20 concerning this situation, the licensee was required by Technical Specification 5.7.2.1 to maintain radiation protection procedures that were consistent with 10 CFR 20. The licensee Procedures RP-143, " Posting of Radiologically Controlled Areas," Revision 9 and TPP 20-2, " Radiation Prc.tection Program," Revision 8, would have also required the area to be postec' as an airborne radioactivity are On December 9,1998, the licensee wrote Corrective Action Request 98-0034 documenting this event. Corrective actions included additional training with the radiation protection staff to include: (1) discussing the above incident; (2) emphasizing the requirements for initial posting and removal of posting for a radiologically area; (3) documenting surveys performed; and (4) updating the survey status board at the entrance to the radiological controlled area. The inspectors verified that on December 10,1998, the above training was conducted with the radiation protection staff. After reviewing the corrective actions taken pertaining to the above event, the inspectors determined that they appeared to be appropriate to resolve the issu Therefore, no response to the above violation is require Conclusions

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Radioactive material containers were labeled, posted, and controlled. All laundry and trash containers were properly maintained. Contaminated areas were posted and clearly identified. Portable radiation protection instrumentation observed in use in the radiological controlled area had been calibrated and source response checked. One violation was identified for the failure to maintain posting of an airborne radioactivity area as required by 10 CFR 20. Based on the corrective actions identified and taken by the licensee, no response to this violation is require .4 Maintainina Occupational Exoosure As Low As is Reasonably Achievable (ALARA) Inspection Scope Radiation protection personnel associated with the ALARA program for the reactor vessel removal project were interviewed. The ALARA planning and monitoring program for the reactor vessel removal project was reviewe _ - _ _ _ _

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-14-l Observations and Findinas i Radiation protection ALARA personnel were appropriately involved with the planning

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and monitoring of the reactor vessel removal project. Tasks were planned using l industry information and past site experience. A review of the radiation work permit Package No. 980016 was completed. Workers were briefed on the ALARA concept and mohods to accomplish their tasks. Mock-ups were used as necessary. The radiation work permit was clearly written providing appropriate radiological controls to the various work activitie .

, Conclusions The ALARA planning program was effectively implemented. Radiation protection  ;

ALARA personnel were appropriately involved with the planning and monitoring of the reactor vessel removal projec Radwaste, Effluents & Environmental Programs (84750) Inspection Scope The condition of the waste systems and effluent monitors was evaluated to verify that l systems were operable and being maintaine .2 Observation and Findinas Control room personnel were interviewed concerning the radioactive effluant system Walkdowns of the gaseous and liquid effluent monitors and radioactive waste storage tanks were conducted. All radioactive effluent monitors were operational, calibrated and properly maintained. Waste storage tanks and associated pumps were in good physical condition with no visible leakage noted. Housekeeping in these areas was good. No problems were identified with the condition, calibration and maintenance of the equipment associated with the radioactive effluent monitoring system l

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Walkdowns of the gaseous and liquid effluent monitors and radioactive waste storage tanks were conducted. All radioactive effluent monitors were operational, calibrated and properly maintained. Waste storage tanks and associated pumps were in good physical condition with no visible leakage noted.

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l ie i -15-6 Independent Spent Fuel Storage Installation (46053,60854)

6.1 Onsite Construction of the First TranStor Concrete Cask a. Insoection Scope .

The inspector reviewed the preparations for the construction of the first TranStor concrete dry storage cask and observed work activities on November 6 and 7,1998 associated with the initial concrete pour of the cask. This included the pour itself and batch testing of the concrete prior to and during pours, b. Observation and Findinas Design requirements for the TranStor concrete cask were detailed in Specification PCC-SP-001," Specification for the Concrete Construction of the TranStor Concrete Cask," Revision 4, dated October,1998. The specification was thorough and referenced the appropriate codes and standards. Guidance was provided regarding construction procedures and practices, material requirements, concrete design, production, testing, and quality assurance. Consideration had also been given to possible changing environmental conditions which could occur during the concrete pourin :

The actual construction of the concrete cask was controlled by an installation / inspection l

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checklist, Form BNFL-007, in Procedure PCC-SP-006,"TranStor Concrete Cask Construction Work Traveler and Inspection Forms," Revision 0, dated October,199 The checklist contained the steps and hold points for cask construction. The inspector reviewed the checklist several times before and during the concrete pour and noted that prerequisites were completed as specified, that steps were being signed off as ]

completed, and that hold points / inspections were being observed by construction personne The testing of the first batch of concrete delivered to the site on November 7,1998, was observed. The tests were performed by qualified and experienced personnel. The licensee rejected the first truck after both slump and air content were determined to be i outside acceptable limits. Actual slump was less than 4 inches. Four to 8 inches was )'

required. Air content was 2.5 percent, with 3 percent to 6 percent required. The second concrete batch tested within the allowed values. Slump was 6.5 inches with an air content of 4 percent. All further batches met slump and air content requirements. One concrete pour was not completed within the required 90-minute time frame from time of manufacture to time of pour. The licensee stopped the pour and retested the concret ,

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After finding the concrete acceptable, the pour was restarted. The licensee's actions were appropriate and demonstrated good attention to detail and procedural adherenc Compression tests for the concrete were performed for the licensee by an outside testing consultant. The acceptable criteria for the compressive strength of the concrete at 28 days was 4,000 pounds / square inch (psi). Test sample results ranged from 7780 psi to 9960 psi, far exceeding the required acceptance criteri _ _ ~._ ___ . _ . _ _ _ _ _ _ _ _ _ -.

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measures had been taken with regard to work performed at the top of and inside the l concrete cask construction area, c. Conclusion l

The licensee completed the construction of the first Transtor concrete cask in accordance with approved procedures. The concrete met the requirements specified for the cask. No discrepancies or concerns were identified with the construction activitie ,

6.2 Dry Cask Storaae Activities i

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a. Inspection Scope Activities and schedules related to the licensee's ISFSI were reviewe b. Observation and Findinas l The licensee had procured a transfer cask, prototype basket, and prototype overpack ,

for the ISFSI. A TranStor concrete storage cask was constructed during this inspectio I Welding equipment had been procured and was being tested. The concrete pad and security system around the pad were in-place. Security system testing was underwa Training for personnel who will be assigned to ISFSI activities was underwa The licensee planned to conduct two dry runs in preparation for loading fuel in dry cask storage at Trojan. The first dry run will encompass the activities related to the transfer station located on the ISFSI pad. This dry run was scheduled for mid January,199 The second dry run willinvolve the activities associated with loading a cask in the fuel building and moving the cask to the ISFSI storage pad. This was scheduled for late March 1999. Current plans were to begin loading fuel into casks in the spring of 1999, assuming completion of the licensing process for the cask design. A total of 34 casks will be used by Trojan, c. Conclusion Plans for loading spent fuel into dry cask storage were aggressively moving forwar Two major walk-throughs of cask loading and unloading procedures were planned for January and March 1999 with possible fuel loading as early as April 199 Follow-up of Open items (92701,92702)

7.1 (Closed) Violation 50-344/9703-01: Failure to Perform Adecuate Surveys: This violation involved the incorrect setting of the low count rate alarm on the CM-11 instrumentation used to perform surveys of material released from the site. The violation was issued as the result of an inspection conducted on July 21-24,1997. The licensee committed to 20 corrective actions in their response, dated November 26,1997, to the violation. Of i

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i this number,11 had been completed prior to issuance of the November 26,1997, lette <

On May 20,1998, a status of the remaining corrective actions was provided to the NRC. Six of the nine remaining corrective actions had been completed. Inspection Report 50-344/98-02, dated July 1,1998, included a review of the corrective actions and found that satisfactory progress was being made. On August 19,1998, the licensee issued a letter to the NRC providing closure of the remaining three corrective action These included purchase of a tool monitor, completion of a resurvey of selected areas of the ISFSI area, and purchase of a pipe / scaffold tube monitor. The tool monitor had ;

been purchased and was operational. The ISFSI area had been surveyed and no '

contamination was found. The pipe / scaffold monitor had been purchased but was not yet operational. The use of the pipe / scaffold monitor was intended to make the survey !

effort for these types of material more automated and quicker. The current process of !

hand surveying the pipe and scaffold material was considered an acceptable process for l material surveys. Even though the new system was not operational, and may be used )

in a much reduced way than originally planned, the current survey methods are  ;

acceptable to the NRC. Therefore, this corrective action is considered complete and the !

corrective actions taken for the violation are considered appropriate. This violation is close )

7.2 (Closed) Follow-up ltem 50-344/9704-01: Corrective Actions-Survevs: The licensee initiated an aggressive effort to determine the extent of contamination outside the radiologically controlled area, implement corrective actions to prevent recurrence, and perform an independent assessment of the radiological protection program. The licensee completed the assessment of the extent of contamination released from the radiologically controlled area and had implemented a new release control program for contaminated material. These programs have been documented as part of the licensee's response to Violations 50-344/9703-01 and 50-344/9705-01. The independent assessment of the radiation protection program was conducted by the licensee's quality assurance department between September 25 and October 30,199 Included as part of the assessment team was an independent consultant with 20 years of health physics experience. The assessment focused on the potential for programmatic deficiencies in the site health physics program with particular attention directed toward the problems associated with the free release program. No major findings were identified during the review of the radiation protection program. This open item is close .3 (Discussed) Violation 50-344/9705-01: Contaminated Material Released from the Radioloaically Controlled Area: As a result of the licensee's radiological survey of material that had been released from the radiologically controlled area, several contaminated items were found. The license committed to a number of corrective actions identified in a March 3,1998 letter to the NRC in response to the violation. The corrective actions have been completed. The licensee was preparing a letter to the NRC for closeout of the violatio .4 LClosed) Follow-up Item 50-344/9802-01: Notification Letter of Acceptance of the New Soent Fuel Pool System: As part of the decommissioning effort, the licensee established a modular spent fuel pool cooling system to isolate the cooling and clean-up systems associated with the spent fuel pool from other systems being dismantled. The new

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system consisted of a different design concept for cooling than the old system. The licensee conducted an acceptance test program for the new system and provided the l results to the NRC in a letter dated August 13,1998. The NRC issued a letter dated .

September 17,1998, finding the modular spent fuel pool system acceptable. This open .

item is closed.-  !

i 8 . Exit Meeting '

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The insoector presented the inspection results to members of licensee management at  !

the conclusion of the inspection on December 10,1998. The licensee acknowledged  !

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the findings presented. The licensee did not identify as proprietary any information provided to, or reviewed by, the inspecto ! >

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. ATTACHMENT PARTIAL LIST OF PERSONS CONTACTED Licensee

' A. Bowman, Radiation Protection Supervisor B. Clark, Decommissioning Supervisor J. Cooper, Emergency Preparedness Engineer L. Dusek, Licensing Engineer K. Hanson, Shift Manager D. Heath, Decommissioning Project Engineer G. Huey, Radiation Protection Technical Support Manager T. Meek, Radiation Protection Manager S. Nichols, Decommissioning Projects M,anager K. Oberloh, Shift Manager S. Schnieder, Plant Operations Manager J. Westvold, Nuclear Regulatory Affairs Manager A. Zacharias, Radiation Protection Specialist Burns and Roe R. Morgan, Burns and Roe State of Oreaon

. A. Bless, Resident inspector, Oregon Office of Energy INSPECTION PROCEDURES USED 36801 Organization, Management, and Cost Controls at Permanently Shutdown Reactors 46053 Structural Concrete Work Observation 60801 Spent Fuel Pool Safety at Permanently Shutdown Reactors 60854 Pre-operational Testing of an ISFSI 71801 Decommissioning Performance and Status Review at Permanently Shutdown Reactors 83750 Occupational Radiation Exposure 84750 Radwaste, Effluents and Environmental Programs 86700 Spent Fuel Pool Activities 92701 Follow-up on Open items 92702 Follow-up on Corrective Actions for Violations and Deviations

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ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-344/9804-01 VIO Failure to Maintain Posting of an Airborne Radioactivity Area Closed l l

50-344/9703-01 VIO Failure to Perform Adequate. Surveys 50-344/9704 01 IFl Corrective Actions - Surveys l 50-344/9802-01 IFl Notification Letter of Acceptance of new Spent Fuel Pool System 50-344/9804 01 VIO Failure to Maintain Posting of an Airborne Radioactivity Area j i

Discussed

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50-344/9705-01 VIO Contaminated Material Released from the RCA  ;

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LIST OF ACRONYMS  !

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ALARA As Low As Reasonable Achievable

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' CAM Continuous Air Monitor ,

CFR Code of Federal Regulations l DAC Derived Air Concentration i ISFSI Independent Spent Fuel Storage installation '

ppm parts per million

. psi pounds per square inch RCA Radiologically Controlled Area j i

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