IR 05000344/1999003
| ML20205T568 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 04/20/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20205T550 | List: |
| References | |
| 50-344-99-03, 50-344-99-3, 72-0017-99-03, 72-17-99-3, NUDOCS 9904270291 | |
| Download: ML20205T568 (21) | |
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i- ENCLOSURE L
U.S. NUCLEAR REGULATORY COMMISSION l REGION IV l
Docket No ;72-17 License No NPF-1; SNM-2509 Report No /99-03;72-17/99-03 l
Licensee: Portland General Electric C Facility: Trojan Nuclear Plant
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i Location: 121 S. W. Salmon Street, TB-17 j Portland, Oregon Dates: March 1-4,15-18, and 22-25,1999 Inspectors: Robert J. Evans, P.E., Health Physicist Nuclear Materia!s inspection Branch Division of Nuclear Materials Safety Lee Thonus, Project Manager Non-Power Reactors and Decommissioning Project Directorate Office of Nuclear Reactor Regulation John E. Whittemore, Senior Reactor Inspector Engineering and Maintenance Branch l
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Division of Reacter Safety Approved By: D. Blair Spitzberg, Ph.D., Chief Fuel Cycle & Decommissioning Branch Division of Nuclear Materials Safety Attachments: 1. Supplementalinformation 2. Trojan RVAIR Project Photographs
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9904270291 990420 PDR ADOCK 05000344 0 PDR L_
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EXECUTIVE SUMMARY l Trojan Nuclear Plant NRC Inspection Report 50-344/99-03; 72-17/99-03 Preoperational testing activities were being conducted on specific components by Portland 4
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General Electric in anticipation of the issuance of a 10 CFR Part 72 license authorizing storage L of spent fuelin an independent Spent Fuel Storage Installation (ISFSI). A full preoperational
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dry run test was planned for late March-early April 1999 to demonstrate the licensee's capability L to load a dry cask with spent reactor fuel. The actual movement of fuel was scheduled to begin L durireg early May 199 During this inspection, the NRC attempted to ascertain whether the licensee was ready to
, perform the dry run test and to perform ISFSI operations. The inspectors concluded that the licensee was ready for the dry run preoperational test, but the licensee was not ready for actual fuel handling operations. A high number of open issues remained to be resolved prior to fuel l handling operations for the licensee, For example, the licensee still had not completed crew l staffing and training, there were still a high number of open problem reports and commitments,
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. and the licensee had not received all equipment for fuel loading (specifically, the first pressurized-water reactor (PWR] basket).
l' During the inspection, the NRC also witnessed the lift of the reactor pressure vessel. The NRC l concluded that the lift was conducted by qualified personnel. Also, the NRC noted that the i licensee had correctly performed tha safety reviews for this important plant decommissioning '
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l Decommissionina Performance
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- Site tours were conducted to observe plant and equipment condition, control room .
l staffing, and housekeeping. Radiological controls were properly installed or utilized, the
. control room was properly staffed, and the licensee was maintaining the spent fuel pool l and its support equipment as stipulated in Technical Specifications and related Defueled Safety Analysis Report (Sections 2.2 and 2.5).
- Housekeeping was mixed. Areas inside of the plant appeared adequa'tely maintained !
for the work in progress, although housekeeping in the ISFSI pad storage building was not at a high level (Section 2.2).
- The inspectors observed the lift of the concrete-filled reactor pressure vessel and reviewed the supporting documentation for the lift. The associated safety reviews were ;
deemed acceptable, but the procedures were not detailed and did not address !
anomaliesiin addition, the review and documentation of changes to these procedures ;
was considered weak by the inspector. However, personnel performing the heavy lift '
activities appeared extremely knowledgeable of the methodology and work being performed (Section 2.3 and 2.4).
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3-I Preooerational Testino of An ISFSI
- The NRC inspectors reviewed the licensee's preparedness for ISFSI operations. The I inspectors noted that the ALARA preplanning activities and quality assurance / quality control oversight of the ISFSI program was strong. A high number of open items / issues remained to be resolved prior to actual fuel loading (Section 3.2).
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which were not resolved during this inspection. This program erea will be reviewed during a future NRC inspection (Section 3.3).
. The licensee held its first +FSI emergency preparedness drill. The licensee identified both strengths and weakrssses during a self-review of the drill. Combined with the licensee's good quality assurance program, the licensee appeared capable of being able to self-critique itself (Section 3.4).
Occupational Radiation Exposure I
- The inspectors performed a review of selected portions of the licensee's radiation protection program to ascertain whether the licensee was in compliance with NRC requirements, and accordingly, maintaining a safe working environment. For calender year 1998, occupational doses were well below the NRC's annual limits. Also, the licensee was controlling the sealed sources in their possession in accordance with Defueled Safety Analysis Report requirements. Overall, the licensee appeared to be maintaining good control over the Trojan facility's radiation protection program (Sections 4.2 and 4.3) .
Solid Radioactive Waste Manaaement and Transportation l
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- The licensee had implemented a waste manifest / shipping paper program that complied l with the requirements specified in Appendix G to 10 CFR Part 20 and the recommendations provided in NUREG/BR-0204 (Section 5.2).
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Spoort Details 1 Plant Status The licensee continued to dismantle and decommission the Trojan facility during this inspection period. The licensee performed the following tasks related to the decommissioning of the reactor pressure vessel during March 1999:
Welding of closure plates over the control rod drive mechanism openings on top j of the reactor vessel,
Load testing of the reactor vessel heavy lift equipment,
- Installation of a jacking frame tower inside of containment to lift the reactor vessel,
- Lifting of the reactor vessel about 10 feet to allow for installation of 2 inch thick shielding around the nozzles, and
- Lifting of the vessel another 55 feet to allow for installation of 5-inch thick shielding around the vesse The licensee plans to lift the shielded reactor pressure vessel, rotate the vessel from the i vertical to the horizontal position, and remove the vessel from containment during April-May 199 Also, the licensee was preparing to move spent fuel from the fuel building for storage at the ISFSt. Sierra Nuclear Corporation's TranStor cask storage system was selected for use at the site, and the licensee was expected to receive a site-specific 10 CFR Part 72 license for this storage method in late March 1999. Preoperational testing of the ISFSI support equipment, including the semi-automated welding, gap flush, and vacuum drying systems, was performed. The licensee planned to perform a dry run test of all ISFSI and fuel handling support systems during late March and early April 1999. The movement of fuel was scheduled to begin during early May 1999 following receipt of the 10 CFR Part 72 license and the first PWR baske The licensee plans to construct and utilize 34 concrete casks for storage of the spent fuel and fuel debris on the ISFSI storage pad. The licensee has completed the ,
fabrication of the first three concrete casks. The casks were being assembled at an ,
onsite construction facilit ,
l The licensee also expanded portions of the ISFSI's concrete pad during March 199 I Additional concrete was installed to widen the corner areas to make cask movement l from the fuel building to the ISFSI storage pad a little easier. Grooves were being cut i into the edges of the concrete to allow for deflation of the air pads if a cask were to be i moved too close to the edges of the concrete pad. The grooves were being installed,in
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part, in response to previous NRC concerns related to accidental movement of a loaded cask off the edge of the concrete pad. Other ISFSI related activities recently completed included installation of a video camera in the cask load pit for fuel verification during fuel loading, installation of a water level gauge in the cask load pit, and continued installation of the ISFSI security support systems.
I 2 Decommissioning Performance (71801,37801) Insoection Scope The purpose of this portion of the inspection was to ascertain whether the licensee was conducting decommissioning activities in accordance with Technical Specifications, Defueled Safety Analysis Report (DSAR), and Decommissioning Plan requirement .2 P! ant Tours Plant tours were performed, in part, to ensure that site equipment, radiological controls, and housekeeping were being effectively controlled in all areas of the plan Radiological controls appeared effective. Boundaries between contaminated /non-contaminated and restricted / unrestricted areas were clearly marked. Postings were in place at all locations. Also, the requirements for entry were clearly marked at each entrance point to a restricted area. Radiological survey meters were located throughout the plant, and these meters appeared to be functioning correctly. No instrument with an out-of-date calibration was identified during the plant tour The inspectors routinely visited the control room. The control room was staffed with at least one qualified individual, and the individuals interviewed appeared knowledgeable about plant conditions. Further, the inspector noted that the operators maintained daily j shift logs.- The' logs were used to document routine and non-routine operational )
activities as well as selected plant parameter Cleanliness and housekeeping efforts were not at a high level in the storage building on the ISFSI pad. The structure contained standing water, and equipment was scattered about the building in a disorganized manner. Wire was being stored in standing water, and the wire exhibited signs of recent rusting. However, housekeeping in the plant's restricted areas was generally adequate for the amount of decommissioning work that was in progress in those areas of the plan .3 Reactor Vessel Lift and Transfer The inspector reviewed the licensee's process for removing the irradiated reactor vessel from the containment. The licensee's methodology, procedures, and concern for industrial and radiation safety were evaluated to determine the effectiveness of the hcensee's control of the heavy loads to be hf ted during the vessel removal proces .
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. l 2. Lift Methodology and Testing The licensee had requested and received NRC approval to remove the reactor vessel with head installed and the vessel internals intact as a single unit, and to transport the vessel on a barge to the final storage location. The alternative was to remove the internals and cut up the reactor vessel for piecemeal shipment to the final storage are The entire vessel with internals in place and the head installed was grouted with low density concrete, and two separate external shielding packages were designed and fabricated for installation. Additionally, the licensee performed an analysis to validate that dropping the reactor vessel during the evolution, at the lowest expected
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temperature, would not induce brittle fracture of the vesse l The licensee's contractor for lifting, downending, and moving the reactor vessel out of I the containment had constructed, installed, and tested a temporary lifting device. The I lifting device could be erected in short and tall tower configurations, depending on the l phase of the lift and transfer. The device consisted of 8 lead screws, operated by hydraulic cylinder jacks, attached to a lifting device which was attached to the vessel by vessel flange studs and nuts. Each lead screw was threaded through a lower load nut engaged by an annular jacking cylinder and an upper capture nut which was turned down on the lead screw against the upper part of the cylinder during lifting to ensure that a failure would result in a minor, if any, drop. The lead screws were segmented in 10-foot sections and a section was removed after it was threaded through the upper .
capture nut. When the cylinder was repositioned to the bottom of stroke, load was I assumed by the upper capture nut while the bottom nut was run down to reengage the cylinder. The speed of lift was about 2-3 feet per hour, j The inspector reviewed documents and interviewed licensee representative regarding l the weight of the package to be moved out of the containment. According to the review the various weights to be considered were:
Vessel, head, internals, and grout = 520 tons 2-inch shielding package = 9.5 tons 5-inch shielding package = 108 tons ,
Maximum expected load to be handled by the device = 820 tons I The inspector verified that the temporary lifting device had been successfully load tested with test weights to a capacity 1050 tons or about 128 percent of the maximum expected loa 'Just prior to the last week of the inspection, an 8 segment,2-inch thick shielding package had been installed around the vessel over the 8 reactor coolant system nozzles.- At the end of the inspection period, the reactor vessel had been raised to accommodate installation of a 5-inch shielding package. The 3,5-inch shielding segments were placed on carts around and under the vessel. It was planned to lower the vessel to the correct elevation to allow the individual shielding segments to be drawn up and installed around the vessel. The inspector obtained a copy of a radiation dose survey conducted pnor to lowering the vessel. Contact readings varied a
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from 150-300 mrom/ hour. Roadogs out to the reactor cavity wall varied between 15 and 50 mrem / hour. After the shielding was in place, these doses were expected to drop at least by a factor of 1 On March 23,1999, an inspector attended a pre job briefing for installation of the 5-inch shielding package for the reactor vessel below the nozzles. The briefing
. provided an overview of the planned evolution and thoroughly emphasized industrial and radiation safety. The inspector found that the pre-job briefing was well-conducted and effectiv The vessel would remain in an upright-elevated configuration until the 5-inch !
shielding package was installed.' During this period, the licensee also planned to fill ,
the gaps between the vessel nozzles and the 2-inch shielding segments with a i steel-filled epoxy compound. The shielding installation commenced on March 24, 1999, and was ongoing at the end of the inspection. Upon completion of the shielding installation, the vassel would be downended onto a cradle, and the temporary lifting device would be reconfigured to the short-tower configuratio Following the downending evolution, the vessel would be slung under the short-tower temporary lifting device and moved out of the containment on the temporary elevated ;
track system. At the end of the track system, the vessel would be lowered onto a i special transporter for transfer to a barge on the rive The inspector noticed that licensee and contractor personnel involved in the transfer of the reactor vessel were highly-skilled and very familiar with the planned process for vessel removal. Up to the end of the inspection period, the equipment and personnelinvolved in the transfer had performed proficiently. Efficient work i
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practices were evident by personnel radiation exposures remaining consistently and significantly less than ALARA program planned exposures.~ )
2.3.2 Procedures The inspector reviewed a copy of Procedure RVAIR 103," Procedure for Lifting, Downending, and Traversing the Reactor Across the TLP and Lowering it to a Trailer," Revision O. As the inspector understood the plan, the initial portion of the lift was to stop at an intermediate level for the installation of a prefabricated 2-inch thick shielding package of 8 pieces around the vessel over the nozzles. Then the lift would continue until conditions were achieved to install a 3-piece,5-inch shielding package around the lower skirt of the vessel. The evolution of installing the 2 inch shield package was not included in the procedure originally provided to the inspecto The inspector requested a copy of the updated procedure. The master procedure was maintained by the vessel lift coordinator and contained 3 temporary changes that were implemented on March 9-10,1999. One of the changes added the previously omitted step and the other two amounted to the addition of informational notes. It appeared that the procedure change had not received licensee QA or supervisory review. The inspector reviewed Procedure TPP 12-1," Nuclear Division Manual u
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8-Procedure Use and Organization," Revision 2, and determined that the vessel lift procedure was not controlled by this process. The inspector held discussions with licensee personnel and also reviewed l Procedures TPP 14-3, " Work Control Process," Revision 11; and TPP 16-1, l " Material / Service Procurement and Control Process," Revision 4. The licensee -
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pointed out that the procedure changes to the vessel lift procedure were performed to meet the requirements of the werk control and procurement control processes. The
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inspector observed and noted to the licensee that changes made to documents in these procedures were subject to a review that was considerably less stringent than l l the review required of a procedure change. The change to the vessel lift procedure l had been approved by the licensee lift coordinator and had received an engineering review. The only documentation of the change and the review was hand written on
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the master document itsel The inspector observed to a licensee representative that Procedure RVAIR 103 was simplistic and did not mention or address anomalies that could occur during the vessel lift or transfer out of the containment. The inspector was provided and !'
reviewed Memorandum RVAIR 200-98," Actions to be Taken as a Result of Heavy Lift Equipment Failures During Removal of the Reactor Vessel from Containment."
The memorandum was from the heavy-lift coordinator to the decommissioning project manager, dated November 11,1998, with a stated purpose to discuss and document responses to possible equipment failures. The document was an analysis that identified seven phases for moving the shielded reactor vessel out of the containment and onto the transporter. For each phase, loss of electrical power as well as plausible structural, mechanical component, and electrical component failures were identified. -Contingencies were provided for the plausible failures. The inspector noted that this document ignored the early step of installing the 2-inch nozzle shielding package just as the procedure had. The document provided two immediate actions in the summary that should be taken for all failures. Theses actions amounted to stopping the lift and placing the load in a safe condition.
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2.4 Evaluation of the Safety Review Prooram An inspector reviewed the licensee's safety evaluation, supporting analyses, and similar documentation related to the RVAIR Plan. The inspector also reviewed recent revisions
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to the modular spent fuel cooling system safety evaluation, an issue which was
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previously identified as an NRC inspection followup item (IFI) (refer to NRC Inspection y Report 50-344/98-202).
The licensee's staff performed the screening process contained in Trojan Plant Procedure TPP 18-1, "10 CFR 50.59 and Other Evaluations," for the RVAIR Plan. The -
l results of the screening process correctly concluded th8t a 10 CFR 50.59 safety evaluation was required. Safety Evaluation SE 96-051 subsequently concluded that the i RVAIR Plan would not create any new types of accidents, increase the severity of any I
analyzed accident or increase the probability of any previously analyzed accidents.
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The RVAIR project includes cutting and capping piping connections, filling the reactor vessel (with internals intact) with low density concrete, raising the reactor vessel, l installing shielding around the vessel, and moving the reactor vessel outside of !
containment to a transporter for shipment to the disposal facility near Richland, Washington. The cutting, capping, and filling of the vessel with concrete has since been successfully completed. The inspector focused on the lifting the reactor vessel and installation of shielding during this inspectio There are no cafety re:ated components remaining in the reactor building with the exception of the fuel transfer tube and associated valving. The activities associated with l the reactor vessel lift, shielding installation, and vessel movement (including the load 1 path) do not take place near the fuel transfer tube and thus should not have an impact on any safety related equipment. The licensee's load drop analysis for the reactor vessel concluded that the vessel would not be subject to brittle failure but would deform i in a ductile, plastic manner. This limits the potential for radioactive releases to 0.155 Curies, which is less than a tenth of previously analyzed releases involving material handling events. (The licensee had analyzed material handling accidents of up to 2.07 Curies of a standardized particulate mixture in the approved Decommissioning Plan.)
An inspector examined the licensee's recent revisions to the Modular Spent Fuel l Cooling System Safety Evaluation SE 98-010. The licensee added information on the limits for radioactive concentrations in the SFP such that they would not exceed the concentrations assumed in the accident analysis. The licensee also added guidance related to freeze protection for the cooling coils. Based on the licensee's revisions to Safety Evaluation SE 98-010, IFl 50-344/98202-01 is close The inspector concluded that the licensee's safety evaluation process for the lifting of
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the reactor vessel and installation of shielding was acceptable. The licensee's revision of the modular spent fuel cooling system safety evaluation resulted in the improvement of this particular safety evaluation thus allowing closure of a previously identified IFl.
l 2.5 Soent Fuel Pool Status Technical Specifications provide operationallimits for the SFP. Further, the Defueled Safety Analysis Report (DSAR) provides additional requirements such as limits for the l
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SFP water chemistry. The inspector confirmed the licensee was maintaining the SFP as stipulated in Technical Specifications and the applicable sections of the DSAR. Proper l operation of the SFP and associated support equipment helps ensure the safe storage j of the reactor fuelin the SFP.
( Technical Specifications 3.1.1 states that the SFP water level shall be greater than or equal to 23 feet over the top of the irradiated fuel assemblies seated in the storage rac The inspector noted on several occasions that the actual water level was about 24 feet, 7 inches. Technical Specifications requires that this parameter be monitored at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by the licensee. The inspector confirmed that this parameter was being monitored by plant operators during each daily shif _
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- 10-Technical Specifications 3.1.2 states that the SFP boron concentration shall be greater than or equal to 2000 parts per million. This parameter is required to be monitored on a weekly basis. The SFP boron concentration was 2181-2186 parts per million during the inspection. The inspector reviewed the licensee's chemistry records and confirmed that the pool had been sampled weekly, and the weekly sample results indicated that the boron concentration had been greater than 2000 parts per million at all times during the previous yea Technical Specifications 3.1.3 states that SFP coolant temperature shall be less than or equal to 140 degrees Fahrenheit. The pool temperature was 78 to 81 degrees during the inspectio DSAR Table 4.3-1, ' Spent Fuel Pool Chemistry Specification and Sampling Schedule,"
lists a number of chemical analyses that are required to be routinely performed on the SFP water. The Table lists the sampling requirements for gross alpha and beta concentrations, tritium, conductivity, chlorides, fluorides, sodium, and suspended solid The inspector confirmed that the licensee had performed the sampling and analyses at the required frequency, and the sample results were well below the limits established in the DSAR for each paramete During plant tours, the operability of the modular spent fuel cooling skid was inspecte The skid appeared to be performing its intended task with all operating parameters, such as pump flow rates, within the expected operational limits. Also, four pool water l makeup sources were available if needed, including service water, fire water, domestic l water, and the " Hale" pum In summary, the licensee was maintaining the SFP in accordance with the requirements specified in Technical Specifications and the DSA .6 Conclusions i
Site tours were conducted to observe plant and equipment condition, control room l staffing, and housekeeping. Radiological contro!s were properly installed or utilized, '
including maintenance of boundaries between the restricted / unrestricted areas, use of postings, and use of personnel monitoring devices. The control room was properly staffed, and plant equipment stillin use appeared to be operable. The licensee was maintaining the SFP and the pool's support equipment as stipulated in Technical Specifications and related DSAR requirements. Housekeeping was mixed. Areas inside of the plant appeared adequately maintained for the work in progress, although housekeeping in the ISFSI storage pad building appeared to need improvemen The inspectors observed the lift of the reactor pressure vessel, and the inspectors reviewed the supporting paperwork for the lift. The associated safety reviews were deemed acceptable, and the licensee's contractor had provided an effective method and appropriate equipment for removing the reactor vessel from the containmen The personnelinvolved were highly-skilled and the equipment had performed well .
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through the initial phases of the vessel transfer. The procedures in use provided the
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process for changing the procedures did not require or result in strenuous review or significant documentation. Despite these observations, the inspectors found that the
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skill and experience of the personnel involved provided adequate preparations for the
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lifting operation l 3 Preoperational Testing of an ISFSI (60854)
l Inspection Scoce The licensee's efforts to develop, implement, and evaluate their preoperational testing activities were reviewed to ensure that the licensee has developed a program to safely load spent fuel from the SFP into a dry cask storage system and to transfer the loaded dry cask to the ISFSt. This review specifically included the licensee's preparedness for the upcoming ISFSI dry run operations, scheduled to be conducted during late March and early April 199 l
! Preparedness for ISFSI Preocerational Testina During this inspection, the licensee's ISFSI programs were reviewed to ascertain whether the licensee was ready to perform the dry run test. The program areas j reviewed included staffing, training, quality assurance / quality control oversight, 1
. procedures, and open items / issues. Overall, the licensee appeared ready for the !
I preoperational testing, but the licensee was not ready for fuel handling operation The licensee had created a staffing plan for ISFSI operations. The licensee planned to I establish four crews. Each crew will consist of twelve individuals. Each crew would be
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led by an ISFSI supervisor who will be a certified fuel handler. Crew members will
' include a technical director, ISFSI loading specialists (equipment operators and l riggers / crane operators), security officers, welders, radiation protection technicians, and quality control specialists. At the time of this inspection, the licensee had staffed and
- trained only two of four crews, and crew training / staffing was still in progress. At least two crews will be fully trained prior to the performance of the dry run test. The licensee expects to have all four crews fully staffed, trained, and certified prior to actual fuel movemen The licensee's quality assurance / quality control groups were actively involved with the l ISFSI projects. The involvement included onsite and offsite audits, inspections, and ;
surveillances. Previous audits performed included design calculations and equipment suppliers. Further, the licensee and their contractors were providing quality oversight during construction of key components, including the PWR baskets and concrete cask Future oversight activities will include the use of hold points in procedures and the g performance of surveillances during fuelloadin I A quality assurance inspector audit was completed during the inspection period. The i audit, -lSFSI Readiness for Fuel Loading," was concluded on March 18,1999. No major L
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! - negative findings were identified, although the auditors noted that the completion of all readiness tasks (closing of open items, approval of draft procedures, completion of i training, receipt and inspection of key components) prior to the start of fuel loading will l be manpower intensive. Strengths were identified in the welding, communications, and
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radiation protection support program areas. The NRC inspector concluded that the
- licensee was dedicated to performing quality work during ISFSI operations, and the t inspectur considered the quality assurance / quality control program area to be a licensee strengt Site procedbres had been drafted for ISFSI operations, although not all procedures had been reviewed, approved, and distributed for site use at the time of this inspection. The licensee planned to complete the review and approval process following receipt of the
- 10 CFR Part 72 license and following use of the draft procedures during preoperational testing. Use of the procedures during preoperational testing will help the licensec verify and validate the completeness and accuracy of the procedure The licensee plans to utilize seven radiation work permits in support of preoperational testing and fuelloading operations. As part of the preplanning process, as low as reasonably achievable (ALARA) reviews were performed to help ensure that occupational doses will be as low as possiblo. The licensee estimated that the fuel loading operations will result in about 24.5 person-rems of exposure. To help reduce the potential exposure, the licensee will take actions that include incorporating " lessons learned" from other licensees from around the country, providing specialized training for potentially high radiation exposure tasks, performing pre-work shift briefings on a routine basis, use of dedicated radiation protection technicians to support fuel movement, and procurement of specialized radiation protection shielding. The inspector considered the licensee's ALARA pre-planning activities to be aggressive. These activities should help reduce or minimize occupational exposures during fuel handling operation During the inspection, the licensee's open issues related to the ISFSI were reviewe The licensee had roughly 140 open items on their commitment tracking list. The licensee also had a number of open items on their ISFSI problem report and corrective action request lists. The overall number of problem reports and corrective action requests indicate that these programs were effectively being used to identify and correct problems, but that a high number of issues were still open. The NRC will review this subject area during a future inspection to ensure that all critical items related to ISFSI operations have been adequately resolved prior to actual fuel handling operation Therefore, this issue has been added to the NRC's preoperational test inspection plan
- checklis In summary, the licensee was not ready for fuel handling operations at the end of this
! inspection period. The licensee still had a number of open issues that needed resolution, including closure of open items, approval of procedures, completion of
. training, and receipt of key components. However, the licensee appeared to have the minimum amount of work completed to perform preoperational testing. The licensee's readiness for ISFSI operations will be reviewed in detail during future NRC inspections.
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I l Area Radiation Monitor Operability Two area radiation monitors (ARMS) serve as criticality alarms in the fuel building spent fuel handling and storage areas. Operability requirements are specified in Section 5.6.1.3.1 of the DSAR. The DSAR requirements include alarm setpoints, surveillance requirements, and contingency actions to take if the ARMS are inoperable.
l If either ARM is inoperable, then the licensee is compelled by the DSAR to take area surveys using portable monitoring equipment on a daily basis. An inspector reviewed the operability of the ARMS to determine whether the monitors were being maintained as stipulated in the DSAR.
! Overall, the ARMS were being maintained in accordance with the requirements of the DSAR, but the inspector questioned whether the ARMS were ready for use during fuel l handling operations. The inspector confirmed that the channel checks, functional tests,
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and calibrations were being performed at the correct intervals. The inspector also confirmed that the alarm setpoints had been established at the values specified in the calibration procedur However, the inspector noted that ARM-12 was partially shielded by an equipment ski If the skid were to remain in its current location during cask movement operations, then the radiation monitor may not be able to perform its intended function because the skid would provide a shield between the radiation source (the cask) and the radiation detector (ARM-12). Also, a corrective maintenance request was outstanding for ARM-12 because the alarm was experiencing spurious alarms. Even though the ARM
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was sporadically malfunctioning, the licensee considered the alarm operable since it passed the daily channel checks. The NRC inspector questioned whether the alarm was truly operable if the alarm was randomly malfunctioning.
l Finally, the inspector noted that the estimated ambient radiation levels during fuel movement may be greater than the ARMS' alarm setpoints, meaning that the alarm would activate each time a cask was transported by the alarms. The licensee is evaluating corrective actions for this issue, and the NRC wiii review the licensee's corrective actions during a future inspectio In summary, this program area will be reviewed in detail during a future inspection to ascertain whether the licensee had adequately resolved these potential issues prior to actual fuel movement. Specifically, the licensee's proposed corrective actions for resolving the ARM operability issue have been added to the NRC's inspection plan checklist for the upcoming ISFSI preoperational inspector inspectio .4 Emeraency Preparedness Drill On March 11,1999, the licensee held its first biennial emergency drill to test the effectiveness of the ISFSI Emergency Plan, implementing procedures, and personnel i training. The accident scenario involved the tip-over of a concrete cask during cask I movement on the ISFSI storage pad. Following the completion of the drill, the licensee concluded that 14 of 16 drill objectives were met. The command and control objective i
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l -14-was not met, while the adequacy of the ISFSI procedures was not completely observe (All ISFSI implementing procedures had not been formally approved at the time of the emergency exercise.)
Two exercise weaknesses were identified: communications between the control room i and the accident scene (communications were incomplete or untimely in several l instances); and the command and control responsibilities at the scene of the accident were not consistent. However, the licensee identified two strengths during the drill. The first strength involved the augmentation and support of the emergency responders (the responders worked well together), while the second strength was related to communications between the onsite responder The licensee concluded that the drill was an overall success, but improvements were needed in certain program areas, including the availability and usefulness of procedures, storage of emergency response supplies within the ISFSI facility, and the role of the access control facility (current location of alarm station and emergency supplies) in a true emergenc l In summary, the licensee held its first ISFSI biennial emergency exercise, and the licensee appeared to be able to independently assess it's strengths and weaknesses in this program area. The licensee's ability to discover and correct these emergency l preparedness program weaknesses will help the licensee be better prepared for a !
potential emergency if one were to occur in the futur .5 Conclusions The NRC inspectors reviewed the licensee's preparedness for ISFSI operations and found that the ALARA preplanning activities and quality assurance / quality control oversight of the ISFSI program were strong. A large number of open items / issues remained to be resolved prior to actual fuel loading. The licensee held its first ISFSI emergency preparedness drill. The licensee identified both strengths and weaknesses during a self review of the drill. Combined with the licensee's good quality assurance program, the licensee appeared capable of being able to self-critique itself. The inspectors identified potential operational problems with the ARMS which were not resolved during this inspection period. The licensee's corrective actions taken to resolve the ARM operability issue will be reviewed during a future NRC inspectio Occupational Radiation Exposure (83750)
4.1 Inspection Scoce Portions of the licensee's radiation protection program were reviewed to ensure that j these programs complied with NRC requilements. In particular, the occupational exposures of site workers were reviewed as well as the licensee's ability to maintain control of sealed radioactive sources.
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i 4.21 Occuoational Fwr-res for 1998 l
10 CFR Part 20 specifies the occupational dose limits for adults, including whole body ,
dose, extremity dose, and dose to an embryo / fetus. The NRC inspector reviewed the licensee's occupational dose records for 1998 to ensure that no individual exceeded the l respective dose limits specified in 10 CFR Part 20.
l l The licensee monitored occupational workers' external exposures using - ;
j- thermoluminescent dosimeters (TLDs) that were being exchanged on a quarterly - i L frequency. During 1998, the licensee monitored 613 individuals. The highest total L effective dose equivalent (TEDE) was 726 millirems, a value that was well below the ;
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NRC's annual limit of 5000 millirems as listed in 10 CFR 20.1201. The highest j exposures were assigned to decommissioning craft personnel, followed by the radiaton -
l protection and maintenance staff personne ;
i TEDEs typically include both extemal and internal doses. The licensee monitored J workers for internal exposures, and the highest internal exposure was about
- 25 millirems, a value that was well below the NRC's ' annual limit of 5000 millirems !
! (10 CFR 20.1201). Since the internal exposures were less than 10 percent of the NRC ,
limit and since the licensee did not anticipate that any individual would actually receive l more than 10 percent of the limit, the licensee did not include internal exposures with !
. the extemal exposures for derivation of the annual TEDEs. Therefore, the assigned l
- TEDE doses were only external doses that were measured by the TLDs. The NRC
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inspector confirmed that the method used by the licensee to determine TEDEs for occupational workers was in compliance with the guidance provided in 10 CFR 20.1502,
" Conditions Requiring Individual Monitoring of Extemal and Intemal Occupational Dose."
!' Finger rings are occasionally used to measure workers' extremity doses. Finger rings were assigned to selected site workers during the third quarter of 1998. The rings were issued to workers that were sorting radioactive debris. The highest measured extremity I dose was 501 millirems. The NRC's extremity dose limit is 50,000 millirems l (10 CFR 20.1201).'
During 1997-1998, one worker was officially designated as a declared pregnant worke In order to limit the potential doses to the fetus / embryo organ, the NRC has established in 10 CFR 20.1208 an occupational dose limit of 500 millirems for the entire gestation period. This particular individual's occupational dose was O millirems during the period of time she was a declared pregnant worke During the first 2 months of 1999, the highest occupational dose was 426 millirem The work projects with the highest potential for occupational dose consist of the cutting,
. packaging, and lifting of the reactor pressure vessel. Other projects with the potential for elevated radiation exposures include waste processing and decommissioning work inside of containment. The licensee continues to monitor worker doses on a routine
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basis, and the licensee does not anticipate that any individual will come close to the
.NRC's annual limits during 199 l L .
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- 16-In summary, the licensee had established an effective program to monitor occupational doses, and occupational workers' doses were well below the NRC limits specified in 10 CFR Part 2 .3 Sealed Sources Inventories and Leak Tests DSAR Section 5.6.3.3.2, " Sealed Sources," states that each sealed source shall be leak tested at least once per 6 months. Also, a physical inventory of all sealed sources shall be performed annually. Finally, the DSAR states that the leak test results as well as the physical inventory records shall be maintained for at least 5 year The inspector reviewed the licensee's sealed source inventory and leak test programs during this inspection. The licensee possessed roughly 400 sealed sources. The sources include those installed in permanent plant equipment as well as radiological survey instrument check sources and radiological standard The inspector confirmed that the licensee was performing a semi-annualleak test on each sealed source. The leak tests were performed during January and July of each calender year. The most recent leak test results from January 1999 revealed that none of the sealed sources were leaking radioactive materials. The most recent test results were less than the DSAR limit (0.005 microcuries of removable contarnination) by a factor of 1 The licensee was performing the sealed source inventory on a semi-annual basis, although the DSAR specifies an annual frequency. The most recent inventory was completed on January 18,1999, and all sources were accounted fo In summary, the licensee was performing the sealed source leak tests and physical inventories in accordance with the DSAR requirements, all sealed sources were accounted for, and the test results reveal that none of the sources were leaking radioactive material .4 Conclusions The inspectors performed a review of selected portions of the licensee's radiation protection program to ascertain whether the licensee was in compliance with NRC requirements, and accordingly, maintaining a safe working environment. For calender year 1998, occupational doses were well below the NRC's annuallimits. Also, the licensee was controlling the sealed sources in their possession in accordance with DSAR requirements. As noted in Section 2.2 of this inspection report, the inspectors observed that the licensee was effectively controlling radiological boundaries, postings, and personnel monitoring. Therefore, the licensee appeared to be maintaining good control over the radiation protection progra _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _
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- 17-5 Solid Radioactive Waste Management & Transportation of Radioactive Materials (86750) Inspection Scooe I
This portion of the inspection attempted to ascertain whether the licensee properly !
processed, packaged, stored, and shipped radioactive materials. In particular, one I waste rnanifest was reviewed in detail and compared to the requirements specified in l 10 CFR Part 2 .2 Waste Manifest Review Appendix G to 10 CFR Pcrt 20 discusses the requirements for transfer of low-level radioactive wastes intended for disposa: at licensed land disposal facilities. Appendix G states that waste generators who transport low-level radioactive wastes intended for disposal at a licerised land disposal facility must prepare a manifest reflecting i information requested on applicable NRC Forms 540, " Uniform Low-Level Radioactive l Waste Manifest (Shipping Paper)" and 541, " Uniform Low-Level Radioactive Waste Manifest (Container and Waste Description)." This regulation went into effect on March 1,199 During the inspection, one waste manifest was randomly selected for comparison with the guidance provided in NUREG/BR-0204, instructions for Completina NRC's Uniform Low-Level Radioactive Waste Manifest. dated April 1995. The waste manifest (No.98-202) consisted of a shipment of three containers of miscellaneous concrete, metal, wood, and trash. Overall, Waste Manifest No.98-202 complied with the recommendations provided in NUREG/BR-0204 and the licensee's implementing procedures. Two minor observations were reported to the licensee for correction l
involving the use of an out-of date procedure checklist and the recording of alpha i
contamination levels that were assumed versus actually measured. Neither of the l observations were safety significant, and the 'icensee stated that corrective actions
- would be taken in both cases.
! Conclusions The licensee had implemented a waste nanifest/ shipping paper program that complied
. with the requirements specified in Appendix G to 10 CFR Pad 20 and the
! recommendations provided in NUREG/BH-020 Followup (92701) ' (Discussed) IFl 50-344/9901-02: Stocking of Equipment ard Supplies As Required by ISFSI Emergency Plan During a previous inspection the NRC observed that the licensee had not staged all emergency equipment and supplies iaentified in the Emergency Plan at the access l
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lr-18-control facility. The licensee did obtain the required supplies for use during their March 11,1999, emergency drill. However, the licenses has since determined that the access control facility may not be the best location for long-term storage of the supplie For example, to transfer the supplies from the Access Control Facility to the ISFSI l during , emergency requires an individual to pass through three security fence The licensee believes that some of the supplies should be relocated from the access control facility to the ISFSI's storage building. This change in storage locations will -
require a formal change to the Emergency Plan. Therefore, this IFl will remain open pending the licensee's final re3olution of where the ISFSI emergency response supplies
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will be locate .2 (Closed) IFl 50-344/98202-01: Modular spent fuel cooling system safety evaluatio This issue is discussed in Section 2.4 of this inspection Repor Exit Meeting Summary j l The inspectors presented the inspection results at the conclusion of the inspection on l March 25,1999. The licensee acknowledged the findings presented. The licenseo stated that some of the contractor's heavy lifting instructions and drawings were proprietary information,
{ otherwise, no other proprietary information was reviewed by the inspector !
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ATTACHMENT 1 l PARTIAL LIST OF PERSONS CONTACTED i
i Licensee J. Allison, Heavy Lift Coordinator / Project Engineer K. Allison, ISFSI Loading Project Manager A. Bowman, Chemistry / Radiation Protection Supervisor C. Brown, Training Specialist K. Cox, ISFSI Manager L. Dusek, Nuclear Regulatory Affairs Manager D. Gildow, Decommissioning Planning Manager G. Huey, Radiation Protection Technical Support Manager J M. Lackey, Engineering & Decommissioning General Manager i T. Meek, Radiation Protection Manager J. Mihelich, Engineering Manager S. Nichols, Decommissioning Projects Manager D. Nordstrom, General Manager S. Quennoz, Trojan Site Executive & Plant General Manager J. Reid, Quality Assurance Specialist S. Schnieder, Operations Manager A. Zacharias, Radiation Protection Specialist State of Oreoon i
A. Bless, Resident inspector, Oregon Office of Energy INSPECTION PROCEDURES USED 37801 Safety Reviews, Design Changes, and Modifications at Permanently Shutdown Reactors 60854 Pre-operational Testing of an ISFSI 71801 Decommissioning Performance & Status Review at Permanently Shutdown Reactors 83750 Occupational Radiation Exposure 86750 Solid Radioactive Waste Management & Transportation of Radioactive Materials 92701 Followup List of Other documents reviewed:
RVAIR 103," Procedure for Lifting, Downending, ard Traversing the Reactor Across the TLP and Lowering it to A Trailer," Revision 0 TPP 121, * Nuclear Division Manual Procedure Use and Organization," Revision 2
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TPP 14-3, " Work Control Process," Revision 11 TPP 14-9, " Control of Heavy Loads," Revision 5 TPP 16-1. " Material / Service Procurement and Control Process," Revision 4 TPP 18-15. " Determining Reportability of Events or Conditions," Revision 5
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i -2-ITEMS OPENED, CLOSED, AND DISCUSSED Opened None Closed 50-344/98202-01 IFl Modular Spent Fuel Cooling System Safety Evaluation Discussed 50-344/9901-02 IFl Emergency Supplies LIST OF ACRONYMS ALARA As Low As Reasonably Achievable ARM Area Radiation Monitor CFR Code of Federal Regulations DSAR- Defueled Safety Analysis Report ISFSI Independent Spent Fuel Stora0e Installation PWR Pressurized Water Reactor RVAIR Reactor Vessel and Internals Removal (Plan)
SFP Spent Fuel Pool
'TEDE Total Effective Dose Equivalent TLD Thermolurrinescent Dosimeter o
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Attachment 2 TROJAN RVAIR PROJECT l
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