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| issue date = 06/26/2006
| issue date = 06/26/2006
| title = Letter to Steven L. Ceccio from Patrick Isaac University of Michigan Ford Nuclear Reactor - Amendment Decommissioning Plan Approval
| title = Letter to Steven L. Ceccio from Patrick Isaac University of Michigan Ford Nuclear Reactor - Amendment Decommissioning Plan Approval
| author name = Isaac P J
| author name = Isaac P
| author affiliation = NRC/NRR/ADRA/DPR/PRTA
| author affiliation = NRC/NRR/ADRA/DPR/PRTA
| addressee name = Ceccio S L
| addressee name = Ceccio S
| addressee affiliation = Univ of Michigan
| addressee affiliation = Univ of Michigan
| docket = 05000002
| docket = 05000002
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| page count = 88
| page count = 88
| project = TAC:MC3707
| project = TAC:MC3707
| stage = Other
| stage = Acceptance Review
}}
}}


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...................................43.0  EVALUATION.........................................................123.1  Decommissioning Alternative..........................................123.1.1  Conclusions..................................................123.2  Controls and Limits on Procedures and Equipment to Protect Occupational and Public Health and Safety..........................................123.2.1  Project Management Structure....................................123.2.1.1  Decommissioning Organization and Responsibilities............123.2.1.2  Key Licensee Positions...................................15 3.2.1.3  Decommissioning Prime Contractor.........................17 3.2.1.4  Safety Review Committee.................................17 3.2.1.5  Conclusions............................................193.2.2  Occupational and Public Health and Safety..........................203.2.2.1  Radiation Protection.....................................20 3.2.2.2  Health Physics Program..................................21 3.2.2.3  Control of Radioactive Materials............................26 3.2.2.4  Dose Estimates.........................................28 3.2.2.5  Radioactive Waste Management...........................293.2.3  Training Program..............................................333.2.3.1  Conclusions............................................363.2.4  General Industrial Safety Program.................................353.2.4.1  Conclusions............................................363.2.5  Radiological Accident Analyses...................................363.2.5.1  Fire..................................................36 3.2.5.2  Pool Leak.............................................393.2.5.3  Tritium-Loaded Heavy-Water Spill...........................393.2.5.4  Conclusions............................................403.3  Decommissioning Activities...........................................403.3.1  Radiological Status of the Facility..................................403.3.1.1  General...............................................403.3.1.2  Principal Radioactive Components..........................403.3.1.3  Sanitary Sewer Lines....................................44 3.3.1.4  Soil beneath the Reactor Building...........................453.3.1.5  Ground Water..........................................453.3.1.6  Radionuclides..........................................453.3.1.7  Conclusions............................................47  
...................................43.0  EVALUATION.........................................................123.1  Decommissioning Alternative..........................................123.1.1  Conclusions..................................................123.2  Controls and Limits on Procedures and Equipment to Protect Occupational and Public Health and Safety..........................................123.2.1  Project Management Structure....................................123.2.1.1  Decommissioning Organization and Responsibilities............123.2.1.2  Key Licensee Positions...................................15 3.2.1.3  Decommissioning Prime Contractor.........................17 3.2.1.4  Safety Review Committee.................................17 3.2.1.5  Conclusions............................................193.2.2  Occupational and Public Health and Safety..........................203.2.2.1  Radiation Protection.....................................20 3.2.2.2  Health Physics Program..................................21 3.2.2.3  Control of Radioactive Materials............................26 3.2.2.4  Dose Estimates.........................................28 3.2.2.5  Radioactive Waste Management...........................293.2.3  Training Program..............................................333.2.3.1  Conclusions............................................363.2.4  General Industrial Safety Program.................................353.2.4.1  Conclusions............................................363.2.5  Radiological Accident Analyses...................................363.2.5.1  Fire..................................................36 3.2.5.2  Pool Leak.............................................393.2.5.3  Tritium-Loaded Heavy-Water Spill...........................393.2.5.4  Conclusions............................................403.3  Decommissioning Activities...........................................403.3.1  Radiological Status of the Facility..................................403.3.1.1  General...............................................403.3.1.2  Principal Radioactive Components..........................403.3.1.3  Sanitary Sewer Lines....................................44 3.3.1.4  Soil beneath the Reactor Building...........................453.3.1.5  Ground Water..........................................453.3.1.6  Radionuclides..........................................453.3.1.7  Conclusions............................................47  
-iv-CONTENTS (Continued)
-iv-CONTENTS (Continued)
Page3.3.2  Radiological Release Criteria.....................................473.3.2.1  Structure Surfaces......................................47 3.3.2.2  Surface Soil and Sediment................................493.3.2.3  Subsurface and Inaccessible Structures......................523.3.2.4  Conclusions............................................533.3.3  Decommissioning Tasks.........................................533.3.3.1  Characterization Surveys.................................53 3.3.3.2  Dismantlement and Decontamination of the Facility.............533.3.3.3  Final Survey and Report..................................573.3.3.4  Conclusions............................................573.3.4  Schedule....................................................573.3.4.1  Conclusions............................................583.3.5  Proposed Final Status Survey Plan................................583.3.5.1  General Survey Approach.................................583.3.5.2  Instrumentation.........................................58 3.3.5.3  Data Quality Objectives...................................58 3.3.5.4  Classifications of Areas by Contamination Potential.............60 3.3.5.5  Identification of Survey Units...............................613.3.5.6  Demonstrating Compliance................................61 3.3.5.7  Background Reference Areas and Materials...................623.3.5.8  Final Status Survey Design................................62 3.3.5.9  Data Assessment.......................................64 3.3.5.10  Final Status Survey Report...............................643.3.5.11  Change Control........................................653.3.5.12  Conclusions...........................................663.4  Estimated Cost.....................................................663.4.1  Conclusions..................................................663.5  Quality Assurance..................................................663.5.1  Overview....................................................66
Page3.3.2  Radiological Release Criteria.....................................473.3.2.1  Structure Surfaces......................................47 3.3.2.2  Surface Soil and Sediment................................493.3.2.3  Subsurface and Inaccessible Structures......................523.3.2.4  Conclusions............................................533.3.3  Decommissioning Tasks.........................................533.3.3.1  Characterization Surveys.................................53 3.3.3.2  Dismantlement and Decontamination of the Facility.............533.3.3.3  Final Survey and Report..................................573.3.3.4  Conclusions............................................573.3.4  Schedule....................................................573.3.4.1  Conclusions............................................583.3.5  Proposed Final Status Survey Plan................................583.3.5.1  General Survey Approach.................................583.3.5.2  Instrumentation.........................................58 3.3.5.3  Data Quality Objectives...................................58 3.3.5.4  Classifications of Areas by Contamination Potential.............60 3.3.5.5  Identification of Survey Units...............................613.3.5.6  Demonstrating Compliance................................61 3.3.5.7  Background Reference Areas and Materials...................623.3.5.8  Final Status Survey Design................................62 3.3.5.9  Data Assessment.......................................64 3.3.5.10  Final Status Survey Report...............................643.3.5.11  Change Control........................................653.3.5.12  Conclusions...........................................663.4  Estimated Cost.....................................................663.4.1  Conclusions..................................................663.5  Quality Assurance..................................................663.5.1  Overview....................................................66 3.5.2 Quality Assurance for Design, Construction, Testing, Modification, and Maintenance..............................................673.5.3  Quality Assurance for Packaging, Preparation for Shipment, and Transportation of Licensed Material................................673.5.4  Quality Assurance for Final Status Survey and Associated Documentation..683.5.4.1  General...............................................683.5.4.2  Organization...........................................683.5.4.3  Written Quality Assurance Program.........................69 3.5.4.4  Training...............................................69 3.5.4.5  Quality Assurance Records................................69 3.5.4.6  Control of Measuring Equipment............................70 3.5.4.7  Audits and Corrective Actions..............................713.5.4.8  Conclusions............................................713.6  Physical Security...................................................713.6.1  Conclusions..................................................71  
 
====3.5.2 Quality====
Assurance for Design, Construction, Testing, Modification, and Maintenance..............................................673.5.3  Quality Assurance for Packaging, Preparation for Shipment, and Transportation of Licensed Material................................673.5.4  Quality Assurance for Final Status Survey and Associated Documentation..683.5.4.1  General...............................................683.5.4.2  Organization...........................................683.5.4.3  Written Quality Assurance Program.........................69 3.5.4.4  Training...............................................69 3.5.4.5  Quality Assurance Records................................69 3.5.4.6  Control of Measuring Equipment............................70 3.5.4.7  Audits and Corrective Actions..............................713.5.4.8  Conclusions............................................713.6  Physical Security...................................................713.6.1  Conclusions..................................................71  
-v-CONTENTS (Continued)
-v-CONTENTS (Continued)
Page4.0  ADDITIONAL LICENSE CONDITIONS......................................724.1  Conclusions.......................................................735.0  TECHNICAL SPECIFICATIONS...........................................735.1  Conclusions.......................................................7
Page4.0  ADDITIONAL LICENSE CONDITIONS......................................724.1  Conclusions.......................................................735.0  TECHNICAL SPECIFICATIONS...........................................735.1  Conclusions.......................................................7
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==2.0  BACKGROUND==
==2.0  BACKGROUND==


===2.1 Regulatory===
2.1 Regulatory BasisThe contents of proposed DPs for research and test reactors must include the following, asspecified in 10 CFR 50.82(b)(4):*the choice of the alternative for decommissioning with a description of activities (see Section 3.1 below)*a description of the controls and limits on procedures and equipment to protectoccupational and public health and safety (see Section 3.2 below)*a description of the planned FSS (see Section 3.3.5 below)
BasisThe contents of proposed DPs for research and test reactors must include the following, asspecified in 10 CFR 50.82(b)(4):*the choice of the alternative for decommissioning with a description of activities (see Section 3.1 below)*a description of the controls and limits on procedures and equipment to protectoccupational and public health and safety (see Section 3.2 below)*a description of the planned FSS (see Section 3.3.5 below)
*an updated cost estimate for the chosen alternative for decommissioning, comparison ofthat estimate with present decommissioning funds set aside, and plan for assuring the availability of adequate funds to complete decommissioning (see Section 3.4 below)*a description of quality assurance provisions, physical security plan provisions, andtechnical specifications (TSs) in place during decommissioning (see Sections 3.5, 3.6, and 5.0 below)  According to 10 CFR 50.82(b)(5), if the DP demonstrates that the decommissioning will beperformed in accordance with the regulations in this chapter and will not be inimical to thecommon defense and security or to the health and safety of the public, and after notice to interested persons, the Commission will approve, by amendment, the plan subject to suchconditions and limitations as it deems appropriate and necessary. The agency based license conditions for this amendment on Appendix 2 to NUREG-1700, Revision 1, "Standard ReviewPlan for Evaluating Nuclear Power License Termination Plans" (Ref. 5). Furthermore, the staff established a license condition in accordance with the requirement of 10 CFR 50.82(b)(5) stating that the approved DP will be a supplement to the safety analysis report (SAR) orequivalent.As specified in 10 CFR 50.82(b)(6), the Commission will terminate the license if it determinesthat the decommissioning was in accordance with the approved DP, and that the FSS andassociated documentation demonstrate that the facility and site are suitable for unrestrictedrelease in accordance with criteria for decommissioning in 10 CFR Part 20, "Standards for Protection Against Radiation," Subpart E, "Radiological Criteria for License Termination."2.2  Site and Facility Description and Operating HistoryThe FNR site and facility is situated on property owned by UM in Ann Arbor, Michigan. TheFNR is located on the North Campus of UM, which is a tract of 900 acres, approximately 1.25 miles northeast of the central business district of Ann Arbor.The reactor building is a windowless, four-story, reinforced concrete building supported by anintegral post-and-beam structure. The 12-inch exterior walls of the reactor building are integral with the footings and foundation mats. The building is approximately 70 feet wide, 68 feet long,and 69 feet high with 55 feet and 46 feet of the structure exposed above grade on the east and west, respectively (Ref. 2).In 1954, the Phoenix Memorial Laboratory (PML) was completed, and construction of the FNRbegan in 1956. The integrated power generated during operation of the FNR was estimated at17,868 megawatt-days between 1957 and 2003 (Ref. 2). The FNR was an open-pool research reactor of the materials testing reactor design. It was a light-water moderated and cooled nonpower reactor with a heterogeneous core composed of aluminum and enriched uranium-235. The reactor was licensed to operate at 2 megawatt thermal (Mwt) power. After the initial startup of the FNR in 1957, reactor operations were permanently placed in safe-shutdown mode on July 3, 2003. Although conjoined with the FNR, PML is not part of the DP. PML is a four-story, reinforcedconcrete building supported by an integral post-and-beam structure that contains offices, wet and dry laboratories, a machine shop, two hot cells, a cobalt irradiator, and various equipment and storage rooms.To assist the licensee's plans to decommission the FNR, the NRC amended License No. R-28on January 29, 2004 (Amendment No. 47), and on May 1, 2006 (Amendment No. 48), to support cessation of reactor operations. Prior to amending the license, the NRC required thelicensee to remove all reactor fuel elements from the FNR and return these licensed materials to the U.S. Department of Energy (DOE). Table 2-1  Profile of UM FNR General Reactor Information:Owner:UMOperator:UMLicensee:UMArchitect/Engineer:Smith, Hinchman & Grylls, Inc.Nuclear Design:Babcock & Wilcox Co.Construction:Jeffress-Dyer, Inc. and Babcock & Wilcox Co.Principal Uses:Training and researchReactor Operation and Authorization:Initial Criticality:September 19, 1957, 04:00Date Secured:July 3, 2003, 15:37 NRC Utilization Facility License #:R-28NRC Facility Docket #:50-2Maximum Power, Steady State, Mwt:2 fthermal Steady State, Water Reflected (nv):3 x 10 13 n cm-2 s-1 peakSpecific Power (kW/kg 235U):382.4Core Power Density, (kW/l):8.5Fuel Material:UAI x , U 3 O 8Uranium Enrichment, %
*an updated cost estimate for the chosen alternative for decommissioning, comparison ofthat estimate with present decommissioning funds set aside, and plan for assuring the availability of adequate funds to complete decommissioning (see Section 3.4 below)*a description of quality assurance provisions, physical security plan provisions, andtechnical specifications (TSs) in place during decommissioning (see Sections 3.5, 3.6, and 5.0 below)  According to 10 CFR 50.82(b)(5), if the DP demonstrates that the decommissioning will beperformed in accordance with the regulations in this chapter and will not be inimical to thecommon defense and security or to the health and safety of the public, and after notice to interested persons, the Commission will approve, by amendment, the plan subject to suchconditions and limitations as it deems appropriate and necessary. The agency based license conditions for this amendment on Appendix 2 to NUREG-1700, Revision 1, "Standard ReviewPlan for Evaluating Nuclear Power License Termination Plans" (Ref. 5). Furthermore, the staff established a license condition in accordance with the requirement of 10 CFR 50.82(b)(5) stating that the approved DP will be a supplement to the safety analysis report (SAR) orequivalent.As specified in 10 CFR 50.82(b)(6), the Commission will terminate the license if it determinesthat the decommissioning was in accordance with the approved DP, and that the FSS andassociated documentation demonstrate that the facility and site are suitable for unrestrictedrelease in accordance with criteria for decommissioning in 10 CFR Part 20, "Standards for Protection Against Radiation," Subpart E, "Radiological Criteria for License Termination."2.2  Site and Facility Description and Operating HistoryThe FNR site and facility is situated on property owned by UM in Ann Arbor, Michigan. TheFNR is located on the North Campus of UM, which is a tract of 900 acres, approximately 1.25 miles northeast of the central business district of Ann Arbor.The reactor building is a windowless, four-story, reinforced concrete building supported by anintegral post-and-beam structure. The 12-inch exterior walls of the reactor building are integral with the footings and foundation mats. The building is approximately 70 feet wide, 68 feet long,and 69 feet high with 55 feet and 46 feet of the structure exposed above grade on the east and west, respectively (Ref. 2).In 1954, the Phoenix Memorial Laboratory (PML) was completed, and construction of the FNRbegan in 1956. The integrated power generated during operation of the FNR was estimated at17,868 megawatt-days between 1957 and 2003 (Ref. 2). The FNR was an open-pool research reactor of the materials testing reactor design. It was a light-water moderated and cooled nonpower reactor with a heterogeneous core composed of aluminum and enriched uranium-235. The reactor was licensed to operate at 2 megawatt thermal (Mwt) power. After the initial startup of the FNR in 1957, reactor operations were permanently placed in safe-shutdown mode on July 3, 2003. Although conjoined with the FNR, PML is not part of the DP. PML is a four-story, reinforcedconcrete building supported by an integral post-and-beam structure that contains offices, wet and dry laboratories, a machine shop, two hot cells, a cobalt irradiator, and various equipment and storage rooms.To assist the licensee's plans to decommission the FNR, the NRC amended License No. R-28on January 29, 2004 (Amendment No. 47), and on May 1, 2006 (Amendment No. 48), to support cessation of reactor operations. Prior to amending the license, the NRC required thelicensee to remove all reactor fuel elements from the FNR and return these licensed materials to the U.S. Department of Energy (DOE). Table 2-1  Profile of UM FNR General Reactor Information:Owner:UMOperator:UMLicensee:UMArchitect/Engineer:Smith, Hinchman & Grylls, Inc.Nuclear Design:Babcock & Wilcox Co.Construction:Jeffress-Dyer, Inc. and Babcock & Wilcox Co.Principal Uses:Training and researchReactor Operation and Authorization:Initial Criticality:September 19, 1957, 04:00Date Secured:July 3, 2003, 15:37 NRC Utilization Facility License #:R-28NRC Facility Docket #:50-2Maximum Power, Steady State, Mwt:2 fthermal Steady State, Water Reflected (nv):3 x 10 13 n cm-2 s-1 peakSpecific Power (kW/kg 235U):382.4Core Power Density, (kW/l):8.5Fuel Material:UAI x , U 3 O 8Uranium Enrichment, %
235U:<20%Fuel Element Geometry:MTR-18 fuel plates (3.25 in. x 2.94 in. x 34.78in.)Element Cladding Material:AluminumElement Cladding Thickness:0.06 in.Core Configuration:35-40 MTR plate-type fuel elementsCore Active Height:24.0 in.No. of Available Fuel Positions:48Coolant:Light WaterModerator:Light WaterReflector:Light water with heavy water on the north face  The following systems continue in operation:*FNR building utility services that are required for facility surveillance and maintenanceunder possession-only status*FNR manually actuated and automated fire alarm systems*FNR security and radiological alarm systems*FNR water demineralization system
235U:<20%Fuel Element Geometry:MTR-18 fuel plates (3.25 in. x 2.94 in. x 34.78in.)Element Cladding Material:AluminumElement Cladding Thickness:0.06 in.Core Configuration:35-40 MTR plate-type fuel elementsCore Active Height:24.0 in.No. of Available Fuel Positions:48Coolant:Light WaterModerator:Light WaterReflector:Light water with heavy water on the north face  The following systems continue in operation:*FNR building utility services that are required for facility surveillance and maintenanceunder possession-only status*FNR manually actuated and automated fire alarm systems*FNR security and radiological alarm systems*FNR water demineralization system 2.3 Scope of the Decommissioning Project The DP lists the various areas, structures, and components that are included in thedecommissioning project. Some of the specific areas include the reactor pool and associated structures and systems, pneumatic t ube system, cooling system, storage ports, building crane,foundation tile, soil under and around the reactor pool, and other impacted interior and exterior building surfaces (see Figures 2-1 to 2-7). The FSS will include the entire FNR facility, such asthe building, systems, and any other areas, as necessary. Residual radioactivity present inthese structures and components will be decontaminated and/or decommissioned to levels thatwill allow for the unrestricted use of this site.
 
===2.3 Scope===
of the Decommissioning Project The DP lists the various areas, structures, and components that are included in thedecommissioning project. Some of the specific areas include the reactor pool and associated structures and systems, pneumatic t ube system, cooling system, storage ports, building crane,foundation tile, soil under and around the reactor pool, and other impacted interior and exterior building surfaces (see Figures 2-1 to 2-7). The FSS will include the entire FNR facility, such asthe building, systems, and any other areas, as necessary. Residual radioactivity present inthese structures and components will be decontaminated and/or decommissioned to levels thatwill allow for the unrestricted use of this site.


Figure 2-1  UM FNR Site Plan  Figure 2-2  FNR Basement  Figure 2-3  FNR First Floor Figure 2-4  FNR Second Floor Figure 2-5  FNR Third Floor  Figure 2-6  FNR Fourth Floor Figure 2-7  East-West Cross Section of Reactor Pool  3.0  EVALUATIONThe NRC staff has reviewed the licensee's proposed actions to decontaminate, dismantle, anddispose of component parts of the FNR, and to perform an FSS. In addition, the staff's review focused on the licensee meeting the regulatory requirements discussed in Section 2.1 aboveand included consideration of the following:*management responsibilities/commitments and personnel qualifications to continuefollowing applicable regulations, regulatory guides, standards, and health and safety plans, including procedures
Figure 2-1  UM FNR Site Plan  Figure 2-2  FNR Basement  Figure 2-3  FNR First Floor Figure 2-4  FNR Second Floor Figure 2-5  FNR Third Floor  Figure 2-6  FNR Fourth Floor Figure 2-7  East-West Cross Section of Reactor Pool  3.0  EVALUATIONThe NRC staff has reviewed the licensee's proposed actions to decontaminate, dismantle, anddispose of component parts of the FNR, and to perform an FSS. In addition, the staff's review focused on the licensee meeting the regulatory requirements discussed in Section 2.1 aboveand included consideration of the following:*management responsibilities/commitments and personnel qualifications to continuefollowing applicable regulations, regulatory guides, standards, and health and safety plans, including procedures
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-refresher training-given annually to all personnel*hazardous waste operations and emergency response-training for personnel engaged inhazardous substance removal or other activities that potentially expose them to hazardoussubstances and health hazards, which satisfies 29 CFR 1910.120, "Hazardous Waste Operations and Emergency Response"*respirator training and fit testing-training, medical qualification, and fit testing for eachperson who wears a tightly fitting respirator that satisfies the requirements of 10 CFR Part 20, Subpart H, and Regulatory Guide 8.15, "Acceptable Programs for Respiratory Protection" (Ref. 12)*Department of Transportation hazardous materials employee training-training as requiredby 49 CFR Part 172, "Hazardous Materials Table, Special Provisions, Hazardous Materials Communications, Emergency Response Information, and Training Requirements,"
-refresher training-given annually to all personnel*hazardous waste operations and emergency response-training for personnel engaged inhazardous substance removal or other activities that potentially expose them to hazardoussubstances and health hazards, which satisfies 29 CFR 1910.120, "Hazardous Waste Operations and Emergency Response"*respirator training and fit testing-training, medical qualification, and fit testing for eachperson who wears a tightly fitting respirator that satisfies the requirements of 10 CFR Part 20, Subpart H, and Regulatory Guide 8.15, "Acceptable Programs for Respiratory Protection" (Ref. 12)*Department of Transportation hazardous materials employee training-training as requiredby 49 CFR Part 172, "Hazardous Materials Table, Special Provisions, Hazardous Materials Communications, Emergency Response Information, and Training Requirements,"
Subpart H, "Training," provided to all personnel involved in the loading, unloading, or handling of hazardous materials, preparing hazardous materials for transportation (including packaging and preparation of manifests), or responsible for the transportation of radioactive materials or operation of a vehicle used to transport hazardous materials (49 CFR 171.8, "Definitions and Abbreviations")*security requirements for offerors and transporters of hazardous materials-training for inthe facility's security plan that satisfies the requirements of 49 CFR Part 172 for allpersonnel involved in the offering of placarded quantities shipments of hazardous materials*hazard communication training-training covering, at minimum, the proper use of materials,the required PPE, and the emergency procedures associated with these materials for all  personnel on the hazardous chemicals in their work area, as required by29 CFR 1910.1200(h), including update training whenever a new physical or health hazard is introduced into their work area*hearing conservation training-training on the effects of noise on hearing and the purposes,advantages, disadvantages, and attenuation of various types of hearing protective devices*permit-required confined space entry training-training for personnel if entry into confinedspaces is to be performed*lockout/tagout training-training for hazardous energy control
Subpart H, "Training," provided to all personnel involved in the loading, unloading, or handling of hazardous materials, preparing hazardous materials for transportation (including packaging and preparation of manifests), or responsible for the transportation of radioactive materials or operation of a vehicle used to transport hazardous materials (49 CFR 171.8, "Definitions and Abbreviations")*security requirements for offerors and transporters of hazardous materials-training for inthe facility's security plan that satisfies the requirements of 49 CFR Part 172 for allpersonnel involved in the offering of placarded quantities shipments of hazardous materials*hazard communication training-training covering, at minimum, the proper use of materials,the required PPE, and the emergency procedures associated with these materials for all  personnel on the hazardous chemicals in their work area, as required by29 CFR 1910.1200(h), including update training whenever a new physical or health hazard is introduced into their work area*hearing conservation training-training on the effects of noise on hearing and the purposes,advantages, disadvantages, and attenuation of various types of hearing protective devices*permit-required confined space entry training-training for personnel if entry into confinedspaces is to be performed*lockout/tagout training-training for hazardous energy control
*trenching and excavation training-training for the purpose of determining the safety andstability of excavations *fire watch training-training on the proper selection, use, and application of extinguishingagents; characteristics and classification of fires*asbestos abatement training-training on requirements, potential health effects, andcontrols for asbestos abatement*torch/plasma arc cutting, welding, and open flame trainings-training in the use of, andunderstanding the reasons for, protective clothing and equipment, including the need for flame-resistant clothing*tailgate training-routine, short training, given usually at the beginning or end of a regularworkforce briefing, intended to provide a brief review of a safety or programmatic topic, which is applicable to current work activities*other specific mandated training-any other training that may be required by the standardsspecific to the Michigan Occupational Safety and Health Act of 1974 (MIOSHA) or applicable standards before initiating work that may fall within the scope of decommissioning3.2.3.1  Conclusions Based on the review of the licensee's training program as described in the DP, the staffconcludes that the licensee's training program is acceptable. The licensee also recognized thatspecific training would be required to reflect the unique hazards associated with decommissioning operations. While the NRC does not regulate nonradiological hazards asspecified in the Atomic Energy Act, the licensee is aware that personnel involved with decommissioning activities would be subject to training requirements administered by other Federal, State, and local government agencies.3.2.4  General Industrial Safety Program The licensee stated that the RSO, with the cooperation of the full project management team,will be responsible for ensuring that the occupational health and safety requirements for projectpersonnel are met, primarily in terms of compliance with the Occupational Safety and Health  Act of 1973 and MIOSHA. Specific responsibilities include establishing training requirementsfor general safe work practices, reviewing plans and procedures to verify adequate coverage of industrial hygiene and safety requirements, conducting periodic inspections of work areas and activities to identify and correct any unsafe conditions and work practices, coordinating industrial hygiene services as required, and advising the Director on industrial hygiene and safety matters and on the results of periodic safety inspections.All personnel working on the FNR decommissioning project will receive health and safetytraining in order to recognize and understand the potential risks to personnel health and safetyassociated with the work at the FNR. The health and safety training also ensures compliance with the applicable regulatory requirements. Personnel will be trained on the plans, procedures,and operation of equipment to conduct work safely on the FNR decommissioning project.The implementation of occupational health and safety requirements for activities involvingpotential hazards that may be encountered during decommissioning will be evaluated thr oughthe use of a JHA. Each JHA will identify all hazards associated with the activity (e.g., fallprotection, hot work, confined space). The licensee will prepare a procedure implementing theJHA that will be subject to the approval requirements discussed in Section 2.4 of the DP. TheJHA allows the project management, project staff, contractor staff, and UM industrial safety personnel (through the RSO or reactor manager) to specify the controls and processes necessary to protect the safety of individual workers, the UM community, and the public. The JHA will act in concert with the RWP, if required, to complete the protection program. Arepresentative of the UM industrial safety staff, the RSO, or the reactor manager will approvethe JHAs. In their absence, the RSO and the reactor manager can delegate this approval authority.3.2.4.1  Conclusions Based on the review of the licensee's proposed industrial safety program as outlined in the DP,the staff concludes the program is acceptable.3.2.5  Radiological Accident Analyses The licensee evaluated radiological accidents that could potentially occur duringdecommissioning of the FNR. This accident analysis considered areas that contain the highestinventories of radioactive material expected to be present during the decommissioning of the FNR. The results of this analysis adequately bounded the radiological impacts that could reasonably occur during decommissioning. As such, a fire, a pool leak, and a tritium-loaded heavy-water spill were the radiological accidents considered to present the highest potentialconsequences.3.2.5.1  Fire The licensee considered the consequences of a fire during decommissioning of the FNR anddid not find them to be significantly different than the consequences of a fire during reactoroperations. The majority of the materials of construction present in the FNR are metals, concrete, or similar noncombustible materials. Upon termination of reactor operation, most of the combustible materials required for reactor operations were removed from the reactor building to further reduce the potential consequences of a fire. The licensee concludes that it is  highly unlikely that a fire would start or that a fire could become intense enough to ignite thesetypes of materials (including other combustible materials such as rags, wipes, and anticontamination clothing), and thus result in the release of radioactive material. The licensee stated that dry radioactive waste is normally collected in metal pails with lidslocated throughout the facility. Once full, the dry waste is normally transferred into 55-gallondrums meeting the strong-tight requirement for shipment to a licensed waste processor. Small quantities of dry radioactive waste requiring special handling or segregation are stored in plastic 5-gallon pails. The licensee stated that this practice limits the volume of dry radioactive waste that could be ignited in a fire event to a few pounds and serves to lower the potential for a fire toconsume additional waste collections. The licensee contends that any fire involving dry radioactive waste would be limited to a few microcuries of radioactivity from radionuclides contained in the list of expected radionuclides (refer to Table 2-4 of the DP).During a fire involving dry radioactive waste, the emission of airborne radioactivity from the FNRexhaust stack would continue unless operator action is taken, or upon automatic closure of the ventilation dampers when the radioactivity levels exceed 1 millirem (mrem) per hour at thebuilding exhaust radiation monitor (required by the TSs). The licensee stated that for the purposes of the evaluation, the ventilation dampers were assumed to remain open, and an exhaust stack dilution factor of 400 and an emission rate of a minimum of 8000 cubic feet per minute up the FNR exhaust stack was assumed, for a duration of 8 hours.Table 3-1 presents the emissions of individual radionuclides that could be released to theenvironment resulting from a fire without exceeding the airborne effluent concentration (AEC) limits for a full year as specified in Table 2 of Appendix B, "Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage," to 10 CFR Part 20. Table 3-1  Quantities of Individual Expected Radionuclides Producing the Emission ofthe AEC during an 8-Hour FireNuclideIndividual Quantity 1Antimony-125 W class130 mCiBismuth-210m D class0.4 mCi Cadmium-109 W class8.7 mCi Carbon-14 (Monoxide)86 Ci Cesium-134 D class8.7 mCi Cesium-137 D class8.7 mCi Cobalt-60 W class8.6 Ci Europium-152 W, All classes390 mCi Europium-154 W, All classes1.3 mCi Iron-55 W class260 mCi Manganese-54 All classes44 mCi Nickel-59 W class430 mCi Nickel-63 W class173 mCi Scandium-46 Y, All classes13 mCi Silver 108m W class17 mCi Silver 110m W class13 mCi Tritium4.3 Ci Zinc-65 Y, All compounds17 mCi 1Activity = AEC x 28,317 cc/ft 3 x 8,000 cfm x 60 min/h x 8 h x 400  Because these quantities of radionuclides are in the mCi range, the licensee stated that theyare significantly greater than the levels expected in any localized, individual containers of dryradioactive waste (i.e., rags, wipes, and anticontamination clothing). During a fire involving these types of wastes in which the ventilation system is secured shortly after the initiation of thefire, the licensee described that exposure would be limited to those individuals who provide initial emergency fire suppression activities. As such, the bounding consequence of the most credible fire event involving an individual container would result in a maximum exposure of 50 mrem to a member of the public. The construction of the reactor building provides three locations where hot gases from a firewould collect. One of these locations is the area above the reactor pool; the other two locations  are just below the third floor on both the east and west sides of the reactor pool. The licenseeconcludes that this inherent design feature aids in the reduction of the concentration ofradioactive materials in the breathing space near the first, second, and third floors of the reactor building, and limits the inhalation of radioactive materials of individuals who provide initial fire suppression activities, those individuals who evacuated the facility, and those individuals whoare required to reenter the reactor building.In the event of a fire, individuals present in the facility may make a reasonable attempt toextinguish the fire using the portable extinguishers provided throughout the facility. If the firecannot be extinguished, the Ann Arbor Fire Department is summoned, as discussed in the FNR Emergency Plan. According to the licensee, it expects only minimal radiological exposures to be incurred by individuals during the short period while attempting to extinguish the fire in the dry radioactive waste or evacuating the area. In addition, fire-fighting personnel responding to a fire potentially involving radioactive materials are trained to use PPE (including adequate respiratory protection, which the licensee concludes would ensure that any internal exposurewould be significantly less that the 50 mrem bounding dose analyzed for a member of the general public).3.2.5.2  Pool Leak In the event of a major leak from the reactor pool, all water loss would be collected by the floordrains or pass through openings in the first floor to the basement of the reactor building. The dimensions of the reactor basement are large enough to allow for the collection of all 50,000 gallons of water from the reactor pool. Loose radioactive contamination in the water's pathway to the reactor basement would be entrained, which the licensee contends should not cause anincrease in the content of radioactive materials in the pool water above the levels experienced while the reactor was operating. The licensee concludes that the resulting levels of radioactive material from the evaporation of the water spread over the basement and first floors would be limited to the tritium contained in the water and would be less than the evaporation ratesexperienced from the surface or the reactor pool while the reactor was operating. The otherradionuclides would remain in the facility. The licensee stated that during normal operation ofthe reactor pool, prior to reactor shutdown, the 240 square feet of the pool's surface was maintained between 90 F and 116 F with an estimated evaporative loss rate of 4 gallons perhour.3.2.5.3  Tritium-Loaded Heavy-Water Spill The consequences of a spill from a 55-gallon drum of tritium-loaded heavy water would be theemission of tritium via the FNR exhaust stack. Taking credit for the FNR exhaust stack dilution factor of 400, and assuming the emission of 8000 cubic feet per minute up the FNR exhaust stack, the licensee calculated the emission of tritium from the facility to be 9.1 mCi per hourbased upon the AEC limit for a full year as specified in Table 2 of Appendix B to 10 CFR Part 20. The heavy-water reflector contains the most concentrated tritium-loaded heavy water. At 217 Ci of tritium (as of April 2004) in an estimated 50 gallons, the licensee calculated the highest estimated concentration of tritium to be 1.1 mCi/milliliter (mL). Given thisconcentration, a spill from this tank would require the evaporation rate to be limited toapproximately 9 mL per hour if the emission were averaged over an entire year. Any spill oftritium-loaded heavy water could be easily flushed to the floor drains for collection in the hot and cold sumps and eventual collection in the retention tanks. Conservatively, 1 week or less would  be needed to clean up the spill or to stop the tritium evaporation. The licensee calculated thatthe emission rate could increase to 473 mCi/hour over 1 week. This equates to an evaporation rate of 0.473 liters (L) of the tritium-loaded heavy water per hour for the entire week of cleanup activities. The licensee concludes that the emission of tritium at a rate of 473 mCi per hour for the 1 week of cleanup would result in a maximum exposure to an individual of 50 mrem.In the event of a spill of the heavy-water reflector while still in the reactor pool, the licenseestated that the dilution by the water in the pool would decrease the concentration of the tritiumin the water source and result in a lower emission rate of tritium from the facility (see LicenseAmendment Nos. 36 and 46).3.2.5.4  Conclusions The licensee analyzed bounding accidents that may occur during the decommissioning project. Based on the NRC staff's review of the information provided by the licensee, the radiologicalconsequences for the types of accidents that may potentially occur during decommissioning ofthe FNR are bounding and within the limits specified in 10 CFR Part 20. 3.3  Decommissioning Activities
*trenching and excavation training-training for the purpose of determining the safety andstability of excavations *fire watch training-training on the proper selection, use, and application of extinguishingagents; characteristics and classification of fires*asbestos abatement training-training on requirements, potential health effects, andcontrols for asbestos abatement*torch/plasma arc cutting, welding, and open flame trainings-training in the use of, andunderstanding the reasons for, protective clothing and equipment, including the need for flame-resistant clothing*tailgate training-routine, short training, given usually at the beginning or end of a regularworkforce briefing, intended to provide a brief review of a safety or programmatic topic, which is applicable to current work activities*other specific mandated training-any other training that may be required by the standardsspecific to the Michigan Occupational Safety and Health Act of 1974 (MIOSHA) or applicable standards before initiating work that may fall within the scope of decommissioning3.2.3.1  Conclusions Based on the review of the licensee's training program as described in the DP, the staffconcludes that the licensee's training program is acceptable. The licensee also recognized thatspecific training would be required to reflect the unique hazards associated with decommissioning operations. While the NRC does not regulate nonradiological hazards asspecified in the Atomic Energy Act, the licensee is aware that personnel involved with decommissioning activities would be subject to training requirements administered by other Federal, State, and local government agencies.3.2.4  General Industrial Safety Program The licensee stated that the RSO, with the cooperation of the full project management team,will be responsible for ensuring that the occupational health and safety requirements for projectpersonnel are met, primarily in terms of compliance with the Occupational Safety and Health  Act of 1973 and MIOSHA. Specific responsibilities include establishing training requirementsfor general safe work practices, reviewing plans and procedures to verify adequate coverage of industrial hygiene and safety requirements, conducting periodic inspections of work areas and activities to identify and correct any unsafe conditions and work practices, coordinating industrial hygiene services as required, and advising the Director on industrial hygiene and safety matters and on the results of periodic safety inspections.All personnel working on the FNR decommissioning project will receive health and safetytraining in order to recognize and understand the potential risks to personnel health and safetyassociated with the work at the FNR. The health and safety training also ensures compliance with the applicable regulatory requirements. Personnel will be trained on the plans, procedures,and operation of equipment to conduct work safely on the FNR decommissioning project.The implementation of occupational health and safety requirements for activities involvingpotential hazards that may be encountered during decommissioning will be evaluated thr oughthe use of a JHA. Each JHA will identify all hazards associated with the activity (e.g., fallprotection, hot work, confined space). The licensee will prepare a procedure implementing theJHA that will be subject to the approval requirements discussed in Section 2.4 of the DP. TheJHA allows the project management, project staff, contractor staff, and UM industrial safety personnel (through the RSO or reactor manager) to specify the controls and processes necessary to protect the safety of individual workers, the UM community, and the public. The JHA will act in concert with the RWP, if required, to complete the protection program. Arepresentative of the UM industrial safety staff, the RSO, or the reactor manager will approvethe JHAs. In their absence, the RSO and the reactor manager can delegate this approval authority.3.2.4.1  Conclusions Based on the review of the licensee's proposed industrial safety program as outlined in the DP,the staff concludes the program is acceptable.3.2.5  Radiological Accident Analyses The licensee evaluated radiological accidents that could potentially occur duringdecommissioning of the FNR. This accident analysis considered areas that contain the highestinventories of radioactive material expected to be present during the decommissioning of the FNR. The results of this analysis adequately bounded the radiological impacts that could reasonably occur during decommissioning. As such, a fire, a pool leak, and a tritium-loaded heavy-water spill were the radiological accidents considered to present the highest potentialconsequences.3.2.5.1  Fire The licensee considered the consequences of a fire during decommissioning of the FNR anddid not find them to be significantly different than the consequences of a fire during reactoroperations. The majority of the materials of construction present in the FNR are metals, concrete, or similar noncombustible materials. Upon termination of reactor operation, most of the combustible materials required for reactor operations were removed from the reactor building to further reduce the potential consequences of a fire. The licensee concludes that it is  highly unlikely that a fire would start or that a fire could become intense enough to ignite thesetypes of materials (including other combustible materials such as rags, wipes, and anticontamination clothing), and thus result in the release of radioactive material. The licensee stated that dry radioactive waste is normally collected in metal pails with lidslocated throughout the facility. Once full, the dry waste is normally transferred into 55-gallondrums meeting the strong-tight requirement for shipment to a licensed waste processor. Small quantities of dry radioactive waste requiring special handling or segregation are stored in plastic 5-gallon pails. The licensee stated that this practice limits the volume of dry radioactive waste that could be ignited in a fire event to a few pounds and serves to lower the potential for a fire toconsume additional waste collections. The licensee contends that any fire involving dry radioactive waste would be limited to a few microcuries of radioactivity from radionuclides contained in the list of expected radionuclides (refer to Table 2-4 of the DP).During a fire involving dry radioactive waste, the emission of airborne radioactivity from the FNRexhaust stack would continue unless operator action is taken, or upon automatic closure of the ventilation dampers when the radioactivity levels exceed 1 millirem (mrem) per hour at thebuilding exhaust radiation monitor (required by the TSs). The licensee stated that for the purposes of the evaluation, the ventilation dampers were assumed to remain open, and an exhaust stack dilution factor of 400 and an emission rate of a minimum of 8000 cubic feet per minute up the FNR exhaust stack was assumed, for a duration of [[estimated NRC review hours::8 hours]].Table 3-1 presents the emissions of individual radionuclides that could be released to theenvironment resulting from a fire without exceeding the airborne effluent concentration (AEC) limits for a full year as specified in Table 2 of Appendix B, "Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage," to 10 CFR Part 20. Table 3-1  Quantities of Individual Expected Radionuclides Producing the Emission ofthe AEC during an 8-Hour FireNuclideIndividual Quantity 1Antimony-125 W class130 mCiBismuth-210m D class0.4 mCi Cadmium-109 W class8.7 mCi Carbon-14 (Monoxide)86 Ci Cesium-134 D class8.7 mCi Cesium-137 D class8.7 mCi Cobalt-60 W class8.6 Ci Europium-152 W, All classes390 mCi Europium-154 W, All classes1.3 mCi Iron-55 W class260 mCi Manganese-54 All classes44 mCi Nickel-59 W class430 mCi Nickel-63 W class173 mCi Scandium-46 Y, All classes13 mCi Silver 108m W class17 mCi Silver 110m W class13 mCi Tritium4.3 Ci Zinc-65 Y, All compounds17 mCi 1Activity = AEC x 28,317 cc/ft 3 x 8,000 cfm x 60 min/h x 8 h x 400  Because these quantities of radionuclides are in the mCi range, the licensee stated that theyare significantly greater than the levels expected in any localized, individual containers of dryradioactive waste (i.e., rags, wipes, and anticontamination clothing). During a fire involving these types of wastes in which the ventilation system is secured shortly after the initiation of thefire, the licensee described that exposure would be limited to those individuals who provide initial emergency fire suppression activities. As such, the bounding consequence of the most credible fire event involving an individual container would result in a maximum exposure of 50 mrem to a member of the public. The construction of the reactor building provides three locations where hot gases from a firewould collect. One of these locations is the area above the reactor pool; the other two locations  are just below the third floor on both the east and west sides of the reactor pool. The licenseeconcludes that this inherent design feature aids in the reduction of the concentration ofradioactive materials in the breathing space near the first, second, and third floors of the reactor building, and limits the inhalation of radioactive materials of individuals who provide initial fire suppression activities, those individuals who evacuated the facility, and those individuals whoare required to reenter the reactor building.In the event of a fire, individuals present in the facility may make a reasonable attempt toextinguish the fire using the portable extinguishers provided throughout the facility. If the firecannot be extinguished, the Ann Arbor Fire Department is summoned, as discussed in the FNR Emergency Plan. According to the licensee, it expects only minimal radiological exposures to be incurred by individuals during the short period while attempting to extinguish the fire in the dry radioactive waste or evacuating the area. In addition, fire-fighting personnel responding to a fire potentially involving radioactive materials are trained to use PPE (including adequate respiratory protection, which the licensee concludes would ensure that any internal exposurewould be significantly less that the 50 mrem bounding dose analyzed for a member of the general public).3.2.5.2  Pool Leak In the event of a major leak from the reactor pool, all water loss would be collected by the floordrains or pass through openings in the first floor to the basement of the reactor building. The dimensions of the reactor basement are large enough to allow for the collection of all 50,000 gallons of water from the reactor pool. Loose radioactive contamination in the water's pathway to the reactor basement would be entrained, which the licensee contends should not cause anincrease in the content of radioactive materials in the pool water above the levels experienced while the reactor was operating. The licensee concludes that the resulting levels of radioactive material from the evaporation of the water spread over the basement and first floors would be limited to the tritium contained in the water and would be less than the evaporation ratesexperienced from the surface or the reactor pool while the reactor was operating. The otherradionuclides would remain in the facility. The licensee stated that during normal operation ofthe reactor pool, prior to reactor shutdown, the 240 square feet of the pool's surface was maintained between 90 F and 116 F with an estimated evaporative loss rate of 4 gallons perhour.3.2.5.3  Tritium-Loaded Heavy-Water Spill The consequences of a spill from a 55-gallon drum of tritium-loaded heavy water would be theemission of tritium via the FNR exhaust stack. Taking credit for the FNR exhaust stack dilution factor of 400, and assuming the emission of 8000 cubic feet per minute up the FNR exhaust stack, the licensee calculated the emission of tritium from the facility to be 9.1 mCi per hourbased upon the AEC limit for a full year as specified in Table 2 of Appendix B to 10 CFR Part 20. The heavy-water reflector contains the most concentrated tritium-loaded heavy water. At 217 Ci of tritium (as of April 2004) in an estimated 50 gallons, the licensee calculated the highest estimated concentration of tritium to be 1.1 mCi/milliliter (mL). Given thisconcentration, a spill from this tank would require the evaporation rate to be limited toapproximately 9 mL per hour if the emission were averaged over an entire year. Any spill oftritium-loaded heavy water could be easily flushed to the floor drains for collection in the hot and cold sumps and eventual collection in the retention tanks. Conservatively, 1 week or less would  be needed to clean up the spill or to stop the tritium evaporation. The licensee calculated thatthe emission rate could increase to 473 mCi/hour over 1 week. This equates to an evaporation rate of 0.473 liters (L) of the tritium-loaded heavy water per hour for the entire week of cleanup activities. The licensee concludes that the emission of tritium at a rate of 473 mCi per hour for the 1 week of cleanup would result in a maximum exposure to an individual of 50 mrem.In the event of a spill of the heavy-water reflector while still in the reactor pool, the licenseestated that the dilution by the water in the pool would decrease the concentration of the tritiumin the water source and result in a lower emission rate of tritium from the facility (see LicenseAmendment Nos. 36 and 46).3.2.5.4  Conclusions The licensee analyzed bounding accidents that may occur during the decommissioning project. Based on the NRC staff's review of the information provided by the licensee, the radiologicalconsequences for the types of accidents that may potentially occur during decommissioning ofthe FNR are bounding and within the limits specified in 10 CFR Part 20. 3.3  Decommissioning Activities 3.3.1 Radiological Status of the Facility 3.3.1.1  General The licensee listed potential causes of radioactive contamination in the reactor building fromnormal operations and routine activities as well as nonroutine occurrences, operations, accidents, and spills. Based on these historical reviews, in addition to characterization surveysof the facility during nearly 50 years of operation of the FNR, the licensee determined thatevents occurred that led to the radiological contamination of the facility. However, the prompt response and cleanup activities initiated by the facility staff limited contamination in areas notexpected to be contaminated by routine operations. Additionally, the licensee's practice ofperiodic monitoring and maintaining contamination action levels between 3 and 10 times background resulted in a limited number of areas where contamination levels are reported to exist above the anticipated release criteria.3.3.1.2  Principal Radioactive Components The information obtained by the licensee indicates that the radioactive portions of the facility areprimarily confined to the reactor internals and reactor pool. The licensee estimated the radioactivity inventory by considering the constituent elements of the material in question and calculating the duration of exposure to the neutron flux and the energies of the incidentneutrons. The licensee is responsible for performing direct measurements during actual removal and/or dismantlement of components. Those data will be used as the basis forspecifying the necessary safety measures and procedures to maintain exposures at ALARA levels during the various dismantlement, removal, decontamination, and waste packaging and storage operations.3.3.1.2.1  Pool Water  The reactor pool and the primary cooling system contain approximately 50,000 gallons of waterrequiring removal. The water was supplied from potable water through filters and demineralizers. The cleanliness of the water was maintained by a system of filters, and H-OHdemineralizers were used as necessary to maintain the conductivity to less than 5 micro-ohm per centimeter. Chemical additions to the water were not required to maintain the pH between 4.5 and 7.5. Table 3-2 lists the levels of radioactivity in the pool water measured in March
 
====3.3.1 Radiological====
Status of the Facility 3.3.1.1  General The licensee listed potential causes of radioactive contamination in the reactor building fromnormal operations and routine activities as well as nonroutine occurrences, operations, accidents, and spills. Based on these historical reviews, in addition to characterization surveysof the facility during nearly 50 years of operation of the FNR, the licensee determined thatevents occurred that led to the radiological contamination of the facility. However, the prompt response and cleanup activities initiated by the facility staff limited contamination in areas notexpected to be contaminated by routine operations. Additionally, the licensee's practice ofperiodic monitoring and maintaining contamination action levels between 3 and 10 times background resulted in a limited number of areas where contamination levels are reported to exist above the anticipated release criteria.3.3.1.2  Principal Radioactive Components The information obtained by the licensee indicates that the radioactive portions of the facility areprimarily confined to the reactor internals and reactor pool. The licensee estimated the radioactivity inventory by considering the constituent elements of the material in question and calculating the duration of exposure to the neutron flux and the energies of the incidentneutrons. The licensee is responsible for performing direct measurements during actual removal and/or dismantlement of components. Those data will be used as the basis forspecifying the necessary safety measures and procedures to maintain exposures at ALARA levels during the various dismantlement, removal, decontamination, and waste packaging and storage operations.3.3.1.2.1  Pool Water  The reactor pool and the primary cooling system contain approximately 50,000 gallons of waterrequiring removal. The water was supplied from potable water through filters and demineralizers. The cleanliness of the water was maintained by a system of filters, and H-OHdemineralizers were used as necessary to maintain the conductivity to less than 5 micro-ohm per centimeter. Chemical additions to the water were not required to maintain the pH between 4.5 and 7.5. Table 3-2 lists the levels of radioactivity in the pool water measured in March


2004.Table 3-2  Radioactivity of the Reactor Pool Water (March 17, 2004)
2004.Table 3-2  Radioactivity of the Reactor Pool Water (March 17, 2004)
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-use a statistical test other than the Sign test or WRS test for evaluation of the FSS
-use a statistical test other than the Sign test or WRS test for evaluation of the FSS
-increase the radioactivity level, relative to the applicable DCGL, at which aninvestigation occurs  -reduce the coverage requirements for scan measurements-decrease an area classification (i.e., impacted to unimpacted, Class 1 to Class 2,Class 2 to Class 3, or Class 1 to Class 3)-increase the Type I decision error
-increase the radioactivity level, relative to the applicable DCGL, at which aninvestigation occurs  -reduce the coverage requirements for scan measurements-decrease an area classification (i.e., impacted to unimpacted, Class 1 to Class 2,Class 2 to Class 3, or Class 1 to Class 3)-increase the Type I decision error
-result in more than a minimal increase in the environmental consequences notpreviously evaluated in the final SAR (as updated)-foreclose the release of the site for possible unrestricted use*The licensee shall submit reports of any characterization surveys performed that were notpart of the license amendment application and shall submit the completed FSS plan for review prior to performing the FSS.The staff finds that the change control criteria proposed by the licensee will adequately facilitatechanges needed to implement the FSS in a manner that ensures both the safety of workers and the public and facilitates timely decommissioning of the FNR.3.3.5.12  Conclusions The staff has reviewed the licensee's DP concerning the planning of the FSS. The staff findsthat the licensee has adequate experience to develop and implement an acceptable MARSSIMFSS. Once the licensee develops the FSS plan, it will present the plan for review and approvalprior to implementation. The NRC staff concludes this aspect of the DP meets the requirementsof 10 CFR 50.82(b)(4)(iii) and is therefore acceptable.3.4  Estimated Cost The licensee stated that decommissioning of the FNR will be accomplished withoutdismantlement of the building. Table 1-1 of the DP presents the detailed estimated cost todecommission the FNR licensed areas. The factors used in these cost estimates were based upon a detailed cost estimate. Using the "High" cost in Table 1-1 of the DP, the licensee estimated that the project will cost up to $9,781,173. Based on the given "High" estimate, theDP states that UM is committed to providing funding for decommissioning of the FNR, in accordance with 10 CFR 50.75(e)(iv).3.4.1  Conclusions The staff has reviewed the licensee's decommissioning cost estimate and finds that the costestimates are consistent with the scope of work covering decommissioning of the FNR. The licensee stated that the UM Regents have specifically approved the expenditure of funds frominvestment proceeds sufficient to cover the "High" cost estimate. The staff concludes that UM is committed to providing acceptable funding for decommissioning of the FNR.3.5  Quality Assurance
-result in more than a minimal increase in the environmental consequences notpreviously evaluated in the final SAR (as updated)-foreclose the release of the site for possible unrestricted use*The licensee shall submit reports of any characterization surveys performed that were notpart of the license amendment application and shall submit the completed FSS plan for review prior to performing the FSS.The staff finds that the change control criteria proposed by the licensee will adequately facilitatechanges needed to implement the FSS in a manner that ensures both the safety of workers and the public and facilitates timely decommissioning of the FNR.3.3.5.12  Conclusions The staff has reviewed the licensee's DP concerning the planning of the FSS. The staff findsthat the licensee has adequate experience to develop and implement an acceptable MARSSIMFSS. Once the licensee develops the FSS plan, it will present the plan for review and approvalprior to implementation. The NRC staff concludes this aspect of the DP meets the requirementsof 10 CFR 50.82(b)(4)(iii) and is therefore acceptable.3.4  Estimated Cost The licensee stated that decommissioning of the FNR will be accomplished withoutdismantlement of the building. Table 1-1 of the DP presents the detailed estimated cost todecommission the FNR licensed areas. The factors used in these cost estimates were based upon a detailed cost estimate. Using the "High" cost in Table 1-1 of the DP, the licensee estimated that the project will cost up to $9,781,173. Based on the given "High" estimate, theDP states that UM is committed to providing funding for decommissioning of the FNR, in accordance with 10 CFR 50.75(e)(iv).3.4.1  Conclusions The staff has reviewed the licensee's decommissioning cost estimate and finds that the costestimates are consistent with the scope of work covering decommissioning of the FNR. The licensee stated that the UM Regents have specifically approved the expenditure of funds frominvestment proceeds sufficient to cover the "High" cost estimate. The staff concludes that UM is committed to providing acceptable funding for decommissioning of the FNR.3.5  Quality Assurance 3.5.1 Overview Section 1.3.4.1 of the DP briefly describes the quality assurance programs used duringdecommissioning, summarized as follows:*A quality assurance program is applied to the design, fabrication, construction, and testing ofstructures, systems, and components of the facility. These quality assurance requirementswould apply to the remediation activities conducted.  *A quality assurance program, which may or may not be the same as the above-mentionedprogram, is applied to the design, purchase, fabrication, handling, shipping, storing, cleaning,assembly, inspection, testing operations, maintenance, repair, and modification of components of packaging used in the transportation of licensed material.*Additional quality assurance requirements are applied to the FSS and associateddocumentation (e.g., characterization information used in the design of the FSS) to ensure that data and the analysis of the data provided to the NRC in the FSS report are accurateand complete. 3.5.2  Quality Assurance for Design, Construction, Testing, Modification, and MaintenanceThe FNR has a quality assurance program, as discussed in Section 1.3.4.2 of the DP, thatmeets the requirement in 10 CFR 50.34, "Contents of Applications; Technical Information," forestablishing and executing a quality assurance program for the design, construction, testing, modification, and maintenance of a research reactor. The descriptions of the managerial and administrative controls will result in a revision to the current quality assurance program. TheFNR will continue to maintain this quality assurance program for the design, construction,testing, modification, and maintenance (including remediation activities) of the reactor.UM will continue to require that all contractors and subcontractors participating in design,construction, testing, modification, and maintenance (including remediation) activities follow the established quality assurance program. Contractors and subcontractors may recommend or request changes to the quality assurance program. UM may or may not make changes to the quality assurance program after review against applicable guidance or standards recommended.
 
====3.5.1 Overview====
Section 1.3.4.1 of the DP briefly describes the quality assurance programs used duringdecommissioning, summarized as follows:*A quality assurance program is applied to the design, fabrication, construction, and testing ofstructures, systems, and components of the facility. These quality assurance requirementswould apply to the remediation activities conducted.  *A quality assurance program, which may or may not be the same as the above-mentionedprogram, is applied to the design, purchase, fabrication, handling, shipping, storing, cleaning,assembly, inspection, testing operations, maintenance, repair, and modification of components of packaging used in the transportation of licensed material.*Additional quality assurance requirements are applied to the FSS and associateddocumentation (e.g., characterization information used in the design of the FSS) to ensure that data and the analysis of the data provided to the NRC in the FSS report are accurateand complete. 3.5.2  Quality Assurance for Design, Construction, Testing, Modification, and MaintenanceThe FNR has a quality assurance program, as discussed in Section 1.3.4.2 of the DP, thatmeets the requirement in 10 CFR 50.34, "Contents of Applications; Technical Information," forestablishing and executing a quality assurance program for the design, construction, testing, modification, and maintenance of a research reactor. The descriptions of the managerial and administrative controls will result in a revision to the current quality assurance program. TheFNR will continue to maintain this quality assurance program for the design, construction,testing, modification, and maintenance (including remediation activities) of the reactor.UM will continue to require that all contractors and subcontractors participating in design,construction, testing, modification, and maintenance (including remediation) activities follow the established quality assurance program. Contractors and subcontractors may recommend or request changes to the quality assurance program. UM may or may not make changes to the quality assurance program after review against applicable guidance or standards recommended.
Changes to the quality assurance program will be approved as discussed in Section 2.4 of theDP.3.5.3Quality Assurance for Packaging, Preparation for Shipment, and Transportation ofLicensed MaterialSubpart H, "Quality Assurance," of 10 CFR Part 71 specifies the requirements for packaging,preparation for shipment, and transportation of licensed material. The managerial and administrative controls the FNR has established to satisfy the requirements of this subpart, described in Section 2.4 of the DP, differ slightly from those previously used. The NRC hasapproved the current FNR quality assurance program as required by 10 CFR 71.101(c). The licensee will follow the existing quality assurance program and maintain it through timelyrenewal, as necessary, to support packaging, preparation for shipment, and transportation of licensed material during remediation activities.UM will continue to require that all contractors and subcontractors participating in packaging,preparation for shipment, and transportation of licensed material follow the approved quality  assurance program. Contractors and subcontractors may recommend or request changes tothe quality assurance program. UM may or may not make changes to the quality assurance program after review against the requirements of 10 CFR Part 71, Subpart H. The licensee willsubmit revisions to the quality assurance program to the NRC for approval as required by10 CFR 71.101(c) prior to implementation and use for the packaging, preparation for shipment,and transportation of licensed materials.UM may elect to use a contractor's or subcontractor's quality assurance program to fulfill therequirements contained in 10 CFR Part 71, Subpart H, after verification that the contractor's orsubcontractor's quality assurance program is acceptable to UM and has been approved by the
Changes to the quality assurance program will be approved as discussed in Section 2.4 of theDP.3.5.3Quality Assurance for Packaging, Preparation for Shipment, and Transportation ofLicensed MaterialSubpart H, "Quality Assurance," of 10 CFR Part 71 specifies the requirements for packaging,preparation for shipment, and transportation of licensed material. The managerial and administrative controls the FNR has established to satisfy the requirements of this subpart, described in Section 2.4 of the DP, differ slightly from those previously used. The NRC hasapproved the current FNR quality assurance program as required by 10 CFR 71.101(c). The licensee will follow the existing quality assurance program and maintain it through timelyrenewal, as necessary, to support packaging, preparation for shipment, and transportation of licensed material during remediation activities.UM will continue to require that all contractors and subcontractors participating in packaging,preparation for shipment, and transportation of licensed material follow the approved quality  assurance program. Contractors and subcontractors may recommend or request changes tothe quality assurance program. UM may or may not make changes to the quality assurance program after review against the requirements of 10 CFR Part 71, Subpart H. The licensee willsubmit revisions to the quality assurance program to the NRC for approval as required by10 CFR 71.101(c) prior to implementation and use for the packaging, preparation for shipment,and transportation of licensed materials.UM may elect to use a contractor's or subcontractor's quality assurance program to fulfill therequirements contained in 10 CFR Part 71, Subpart H, after verification that the contractor's orsubcontractor's quality assurance program is acceptable to UM and has been approved by the


Revision as of 19:12, 13 July 2019

Letter to Steven L. Ceccio from Patrick Isaac University of Michigan Ford Nuclear Reactor - Amendment Decommissioning Plan Approval
ML061220260
Person / Time
Site: University of Michigan
Issue date: 06/26/2006
From: Isaac P
NRC/NRR/ADRA/DPR/PRTA
To: Ceccio S
University of Michigan
Issac P, NRR/DRIP/REXB, 415-1019
References
TAC MC3707
Download: ML061220260 (88)
v • d • e


Text

June 26, 2006Mr. Steven L. Ceccio, DirectorPhoenix Memorial Laboratory 2301 Bonisteel Boulevard University of Michigan Ann Arbor, MI 48109

SUBJECT:

UNIVERSITY OF MICHIGAN FORD NUCLEAR REACTOR-AMENDMENT RE: DECOMMISSIONING PLAN APPROVAL (TAC NO. MC3707)

Dear Mr. Ceccio:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 50to Facility Operating License No. R-28 for the University of Michigan Ford Nuclear Reactor,Docket No. 50-2.The amendment approves the decommissioning plan (DP) for the Ford Nuclear Reactor inresponse to your application of June 18, 2004, as supplemented on June 23, 2004, January 5,and January 10, 2006. The amendment authorizes the inclusion of the approved DP as asupplement to the safety analysis report pursuant to Title 10, Section 50.82(b)(5), of the Codeof Federal Regulations (10 CFR 50.82(b)(5)). In addition, in accordance with10 CFR 50.82(b)(5), the NRC staff has added license conditions to Facility Operating LicenseNo. R-28 deemed appropriate and necessary for approval of the DP.We have also enclosed a copy of the safety evaluation supporting Amendment No. 50.Sincerely,/RA/Patrick Isaac, Project ManagerResearch and Test Reactors Branch Division of Policy and Rulemaking Office of Nuclear Reactor RegulationDocket No. 50-02

Enclosures:

1. Amendment No. 50
2. Safety Evaluationcc w/enclosures: See next page University of MichiganDocket No. 50-02 cc:

Special Assistant to the GovernorOffice of the Governor Room 1-State Capitol Lansing, MI 48909Mr. C.W. BeckerPhoenix Memorial Laboratory 2301 Bonisteel Boulevard University of Michigan Ann Arbor, MI 48109Michigan Department of Environmental QualityWaste and Hazardous Materials Division Hazardous Waste and Radiological Protection Section Nuclear Facilities Unit 525 West Allegan Street P.O. Box 30241 Lansing, MI 48909-7741Test, Research, and TrainingReactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611 June 26, 2006Mr. Steven L. Ceccio, Director Phoenix Memorial Laboratory 2301 Bonisteel Boulevard University of Michigan Ann Arbor, MI 48109

SUBJECT:

UNIVERSITY OF MICHIGAN FORD NUCLEAR REACTOR-AMENDMENT RE: DECOMMISSIONING PLAN APPROVAL (TAC NO. MC3707)

Dear Mr. Ceccio:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 50to Facility Operating License No. R-28 for the University of Michigan Ford Nuclear Reactor,Docket No. 50-2.The amendment approves the decommissioning plan (DP) for the Ford Nuclear Reactor inresponse to your application of June 18, 2004, as supplemented on June 23, 2004, January 5,and January 10, 2006. The amendment authorizes the inclusion of the approved DP as asupplement to the safety analysis report pursuant to Title 10, Section 50.82(b)(5), of the Codeof Federal Regulations (10 CFR 50.82(b)(5)). In addition, in accordance with10 CFR 50.82(b)(5), the NRC staff has added license conditions to Facility Operating LicenseNo. R-28 deemed appropriate and necessary for approval of the DP.We have also enclosed a copy of the safety evaluation supporting Amendment No. 50.Sincerely,/RA/Patrick Isaac, Project ManagerResearch and Test Reactors Branch Division of Policy and Rulemaking Office of Nuclear Reactor RegulationDocket No. 50-02

Enclosures:

1. Amendment No. 50
2. Safety Evaluationcc w/enclosures: See next pageDISTRIBUTION
PUBLICPRT r/fJQuichochoPIsaacCBassett EHyltonMMendoncaAAdamsOGCDHArrison TDragounKWittDHughesWSchusterMVoth GHill (2) (T5-C3)BThomasADAMS ACCESSION NO: ML061220260OFFICETechEd PRT:LAPRT:RIPRT:PMOGCPRT:BCNAMEPIsaac forEHylton:tls*PIsaac for TDragoun*PIsaac*HWedewer*BThomas:tls*DATE5/11/065/15/065/15/065/16/066/2/066/26/06OFFICIAL RECORD COPY UNIVERSITY OF MICHIGANDOCKET NO. 50-02AMENDMENT TO FACILITY OPERATING LICENSEAmendment No. 50License No. R-28 1.The U.S. Nuclear Regulatory Commission (the Commission) has found that:A.The application filed by the University of Michigan (the licensee), dated June 18, 2004,and as supplemented on June 23, 2004, January 5, and January 10, 2006, complieswith the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as stated in Title 10, Chapter 1, of theCode of Federal Regulations (10 CFR Chapter 1);B.The facility will be possessed and decommissioned in conformity with the application,the provisions of the Act, and the rules and regulations of the Commission;C.There is reasonable assurance (i) that the activities authorized by this amendment canbe conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the rules and regulations of theCommission;D.The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public;E.This amendment is issued in accordance with 10 CFR Part 51 of the regulations of theCommission and all applicable requirements have been satisfied; and2.Accordingly, the license is amended by changes to the following paragraph which is herebyamended to read as follows:1.B.The facility will be possessed and decommissioned in conformity with theapplication, the provisions of the Act, and the rules and regulations of the Commission;3.Accordingly, the license is amended by changes to paragraph 2.C.3 to Facility OperatingLicense No. R-28 which hereby reads as follows:2.C.3.Decommissioninga.The license is amended to approve the decommissioning plan described in the licensee's application dated June 23, 2004, as supplemented on January 5, 2006,and January 10, 2006, and authorizes inclusion of the decommissioning plan as a supplement to the safety analysis report pursuant to 10 CFR 50.82(b)(5).b.The licensee may make changes to the decommissioning plan without priorapproval provided the proposed changes do not:(i)Require Commission approval pursuant to 10 CFR 50.59; (ii)Use a statistical test other than the Sign test or Wilcoxon Rank Sum test forevaluation of the final status survey;(iii)Increase the radioactivity level, relative to the applicable derivedconcentration guideline level, at which an investigation occurs;(iv) Reduce the coverage requirements for scan measurements; (v)Decrease an area classification (i.e., impacted to unimpacted; Class 1 toClass 2; Class 2 to Class 3; or Class 1 to Class 3);(vi) Increase the Type I decision error; (vii) Result in more than a minimal increase in the environment consequences notpreviously evaluated in the final safety analysis report (as updated);(viii)Foreclose the release of the site for possible unrestricted use.

.c.The licensee shall submit reports of all characterization surveys performed that werenot part of the license amendment application and shall submit the completed finalstatus survey plan for review prior to performing the final status survey.4.This license amendment is effective as of the date of its issuance.FOR THE U.S. NUCLEAR REGULATORY COMMISSION/RA/

Brian Thomas, Branch ChiefResearch and Test Reactors Branch Division of Policy and Rulemaking Office of Nuclear Reactor RegulationDate of Issuance:June 26, 2006 SAFETY EVALUATION RELATED TO THE DECOMMISSIONING OFTHE UNIVERSITY OF MICHIGAN FORD NUCLEAR REACTORUNIVERSITY OF MICHIGANJune 2006Office of Nuclear Reactor RegulationDivision of Regulatory Improvement ProgramsOperating Reactor Improvements Program

-ii-ABSTRACTThis safety evaluation summarizes the findings of a technical review conducted by the staff ofthe U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation. Thestaff conducted this review in response to an application filed by the University of Michigan (UM or the licensee) for approval of the decommissioning plan (DP) for the Ford Nuclear Reactor (FNR). The FNR is located on the UM campus in Ann Arbor, Michigan. On the basis of this review, the staff concludes that UM can safely dismantle the FNR and dispose of the component parts in accordance with their DP, as amended, and the NRC's rules andregulations.

-iii-CONTENTS PageABSTRACT................................................................ii

1.0 INTRODUCTION

........................................................

12.0 BACKGROUND

.........................................................12.1 Regulatory Basis

....................................................12.2 Site and Facility Description and Operating History

..........................22.3 Scope of the Decommissioning Project

...................................43.0 EVALUATION.........................................................123.1 Decommissioning Alternative..........................................123.1.1 Conclusions..................................................123.2 Controls and Limits on Procedures and Equipment to Protect Occupational and Public Health and Safety..........................................123.2.1 Project Management Structure....................................123.2.1.1 Decommissioning Organization and Responsibilities............123.2.1.2 Key Licensee Positions...................................15 3.2.1.3 Decommissioning Prime Contractor.........................17 3.2.1.4 Safety Review Committee.................................17 3.2.1.5 Conclusions............................................193.2.2 Occupational and Public Health and Safety..........................203.2.2.1 Radiation Protection.....................................20 3.2.2.2 Health Physics Program..................................21 3.2.2.3 Control of Radioactive Materials............................26 3.2.2.4 Dose Estimates.........................................28 3.2.2.5 Radioactive Waste Management...........................293.2.3 Training Program..............................................333.2.3.1 Conclusions............................................363.2.4 General Industrial Safety Program.................................353.2.4.1 Conclusions............................................363.2.5 Radiological Accident Analyses...................................363.2.5.1 Fire..................................................36 3.2.5.2 Pool Leak.............................................393.2.5.3 Tritium-Loaded Heavy-Water Spill...........................393.2.5.4 Conclusions............................................403.3 Decommissioning Activities...........................................403.3.1 Radiological Status of the Facility..................................403.3.1.1 General...............................................403.3.1.2 Principal Radioactive Components..........................403.3.1.3 Sanitary Sewer Lines....................................44 3.3.1.4 Soil beneath the Reactor Building...........................453.3.1.5 Ground Water..........................................453.3.1.6 Radionuclides..........................................453.3.1.7 Conclusions............................................47

-iv-CONTENTS (Continued)

Page3.3.2 Radiological Release Criteria.....................................473.3.2.1 Structure Surfaces......................................47 3.3.2.2 Surface Soil and Sediment................................493.3.2.3 Subsurface and Inaccessible Structures......................523.3.2.4 Conclusions............................................533.3.3 Decommissioning Tasks.........................................533.3.3.1 Characterization Surveys.................................53 3.3.3.2 Dismantlement and Decontamination of the Facility.............533.3.3.3 Final Survey and Report..................................573.3.3.4 Conclusions............................................573.3.4 Schedule....................................................573.3.4.1 Conclusions............................................583.3.5 Proposed Final Status Survey Plan................................583.3.5.1 General Survey Approach.................................583.3.5.2 Instrumentation.........................................58 3.3.5.3 Data Quality Objectives...................................58 3.3.5.4 Classifications of Areas by Contamination Potential.............60 3.3.5.5 Identification of Survey Units...............................613.3.5.6 Demonstrating Compliance................................61 3.3.5.7 Background Reference Areas and Materials...................623.3.5.8 Final Status Survey Design................................62 3.3.5.9 Data Assessment.......................................64 3.3.5.10 Final Status Survey Report...............................643.3.5.11 Change Control........................................653.3.5.12 Conclusions...........................................663.4 Estimated Cost.....................................................663.4.1 Conclusions..................................................663.5 Quality Assurance..................................................663.5.1 Overview....................................................66 3.5.2 Quality Assurance for Design, Construction, Testing, Modification, and Maintenance..............................................673.5.3 Quality Assurance for Packaging, Preparation for Shipment, and Transportation of Licensed Material................................673.5.4 Quality Assurance for Final Status Survey and Associated Documentation..683.5.4.1 General...............................................683.5.4.2 Organization...........................................683.5.4.3 Written Quality Assurance Program.........................69 3.5.4.4 Training...............................................69 3.5.4.5 Quality Assurance Records................................69 3.5.4.6 Control of Measuring Equipment............................70 3.5.4.7 Audits and Corrective Actions..............................713.5.4.8 Conclusions............................................713.6 Physical Security...................................................713.6.1 Conclusions..................................................71

-v-CONTENTS (Continued)

Page4.0 ADDITIONAL LICENSE CONDITIONS......................................724.1 Conclusions.......................................................735.0 TECHNICAL SPECIFICATIONS...........................................735.1 Conclusions.......................................................7

36.0 ENVIRONMENTAL CONSIDERATION

......................................73

7.0 CONCLUSION

S........................................................73 ABBREVIATIONS..........................................................75REFERENCES............................................................77LIST OF FIGURESFigure 2-1 UM FNR Site Plan.................................................5Figure 2-2 FNR Basement....................................................6 Figure 2-3 FNR First Floor....................................................7 Figure 2-4 FNR Second Floor

.................................................8Figure 2-5 FNR Third Floor.................................................. 9 Figure 2-6 FNR Fourth Floor.................................................10 Figure 2-7 East-West Cross Section of Reactor Pool..............................11Figure 3-1 Organization Chart for the FNR Decommissioning Project..................14Figure 3-2 Radiation Levels (R/hr) on the Reactor Grid Plate (April 2004)..............42LIST OF TABLESTable 2-1 Profile of UM FNR..................................................3Table 3-1 Quantities of Individual Expected Radionuclides Producing the Emission of the AEC during an 8-Hour Fire..........................................38 Table 3-2 Radioactivity of the Reactor Pool Water (March 17, 2004)..................41Table 3-3 Estimated Material Volumes for the Thermal Column......................44 Table 3-4 List of Potential Radionuclides........................................46Table 3-5 Acceptable License Termination Screening Values of Common Radionuclides for Structure Surfaces..............................................48 Table 3-6 Acceptable License Termination Screening Values of Common Radionuclides for Surface Soil......................................................50 Table 3-7 Instrumentation for FNR Radiological Surveys...........................59

1.0 INTRODUCTION

By letter dated June 18, 2004, the University of Michigan (UM or the licensee) (Ref. 1)submitted a license amendment request to the U.S. Nuclear Regulatory Commission (NRC) forapproval of its decommissioning plan (DP), Revision 00, dated June 23, 2004 (Ref. 2), and authorization to dismantle and dispose of component parts of the Ford Nuclear Reactor (FNR).

Subsequently, the licensee submitted Revision 1 to the DP (Ref. 3), dated January 5, 2006. On January 10, 2006 (Ref. 4), the licensee submitted additional detail concerning site characterization.The licensee selected the DECON option as the decommissioning alternative. This option willconsist of decontamination and removal of equipment and material containing residual radioactivity from the site to levels allowing for unrestricted release as specified in Title 10, Section 20.1402 of the Code of Federal Regulations (10 CFR 20.1402). In the licenseamendment request, UM described its plan for developing and implementing the final status survey (FSS) plan to verify and document that the decommissioned areas and structures meet the requirements of release for unrestricted use. Upon completion of decommissioning-related activities and the FSS, UM will submit the necessary documentation for review and approval bythe NRC to support license termination.The NRC published a "Notice and Solicitation of Comments Pursuant to 10 CFR 20.1405 and10 CFR 50.82(b)(5) Concerning Proposed Action to Decommission the University of Michigan Ford Nuclear Reactor (FNR)" in the Federal Register on September 8, 2004 (69 FR54326-54327), and in The Ann Arbor News on September 9, 2004. The agency did not receive any comments.

2.0 BACKGROUND

2.1 Regulatory BasisThe contents of proposed DPs for research and test reactors must include the following, asspecified in 10 CFR 50.82(b)(4):*the choice of the alternative for decommissioning with a description of activities (see Section 3.1 below)*a description of the controls and limits on procedures and equipment to protectoccupational and public health and safety (see Section 3.2 below)*a description of the planned FSS (see Section 3.3.5 below)

  • an updated cost estimate for the chosen alternative for decommissioning, comparison ofthat estimate with present decommissioning funds set aside, and plan for assuring the availability of adequate funds to complete decommissioning (see Section 3.4 below)*a description of quality assurance provisions, physical security plan provisions, andtechnical specifications (TSs) in place during decommissioning (see Sections 3.5, 3.6, and 5.0 below) According to 10 CFR 50.82(b)(5), if the DP demonstrates that the decommissioning will beperformed in accordance with the regulations in this chapter and will not be inimical to thecommon defense and security or to the health and safety of the public, and after notice to interested persons, the Commission will approve, by amendment, the plan subject to suchconditions and limitations as it deems appropriate and necessary. The agency based license conditions for this amendment on Appendix 2 to NUREG-1700, Revision 1, "Standard ReviewPlan for Evaluating Nuclear Power License Termination Plans" (Ref. 5). Furthermore, the staff established a license condition in accordance with the requirement of 10 CFR 50.82(b)(5) stating that the approved DP will be a supplement to the safety analysis report (SAR) orequivalent.As specified in 10 CFR 50.82(b)(6), the Commission will terminate the license if it determinesthat the decommissioning was in accordance with the approved DP, and that the FSS andassociated documentation demonstrate that the facility and site are suitable for unrestrictedrelease in accordance with criteria for decommissioning in 10 CFR Part 20, "Standards for Protection Against Radiation," Subpart E, "Radiological Criteria for License Termination."2.2 Site and Facility Description and Operating HistoryThe FNR site and facility is situated on property owned by UM in Ann Arbor, Michigan. TheFNR is located on the North Campus of UM, which is a tract of 900 acres, approximately 1.25 miles northeast of the central business district of Ann Arbor.The reactor building is a windowless, four-story, reinforced concrete building supported by anintegral post-and-beam structure. The 12-inch exterior walls of the reactor building are integral with the footings and foundation mats. The building is approximately 70 feet wide, 68 feet long,and 69 feet high with 55 feet and 46 feet of the structure exposed above grade on the east and west, respectively (Ref. 2).In 1954, the Phoenix Memorial Laboratory (PML) was completed, and construction of the FNRbegan in 1956. The integrated power generated during operation of the FNR was estimated at17,868 megawatt-days between 1957 and 2003 (Ref. 2). The FNR was an open-pool research reactor of the materials testing reactor design. It was a light-water moderated and cooled nonpower reactor with a heterogeneous core composed of aluminum and enriched uranium-235. The reactor was licensed to operate at 2 megawatt thermal (Mwt) power. After the initial startup of the FNR in 1957, reactor operations were permanently placed in safe-shutdown mode on July 3, 2003. Although conjoined with the FNR, PML is not part of the DP. PML is a four-story, reinforcedconcrete building supported by an integral post-and-beam structure that contains offices, wet and dry laboratories, a machine shop, two hot cells, a cobalt irradiator, and various equipment and storage rooms.To assist the licensee's plans to decommission the FNR, the NRC amended License No. R-28on January 29, 2004 (Amendment No. 47), and on May 1, 2006 (Amendment No. 48), to support cessation of reactor operations. Prior to amending the license, the NRC required thelicensee to remove all reactor fuel elements from the FNR and return these licensed materials to the U.S. Department of Energy (DOE). Table 2-1 Profile of UM FNR General Reactor Information:Owner:UMOperator:UMLicensee:UMArchitect/Engineer:Smith, Hinchman & Grylls, Inc.Nuclear Design:Babcock & Wilcox Co.Construction:Jeffress-Dyer, Inc. and Babcock & Wilcox Co.Principal Uses:Training and researchReactor Operation and Authorization:Initial Criticality:September 19, 1957, 04:00Date Secured:July 3, 2003, 15:37 NRC Utilization Facility License #:R-28NRC Facility Docket #:50-2Maximum Power, Steady State, Mwt:2 fthermal Steady State, Water Reflected (nv):3 x 10 13 n cm-2 s-1 peakSpecific Power (kW/kg 235U):382.4Core Power Density, (kW/l):8.5Fuel Material:UAI x , U 3 O 8Uranium Enrichment, %

235U:<20%Fuel Element Geometry:MTR-18 fuel plates (3.25 in. x 2.94 in. x 34.78in.)Element Cladding Material:AluminumElement Cladding Thickness:0.06 in.Core Configuration:35-40 MTR plate-type fuel elementsCore Active Height:24.0 in.No. of Available Fuel Positions:48Coolant:Light WaterModerator:Light WaterReflector:Light water with heavy water on the north face The following systems continue in operation:*FNR building utility services that are required for facility surveillance and maintenanceunder possession-only status*FNR manually actuated and automated fire alarm systems*FNR security and radiological alarm systems*FNR water demineralization system 2.3 Scope of the Decommissioning Project The DP lists the various areas, structures, and components that are included in thedecommissioning project. Some of the specific areas include the reactor pool and associated structures and systems, pneumatic t ube system, cooling system, storage ports, building crane,foundation tile, soil under and around the reactor pool, and other impacted interior and exterior building surfaces (see Figures 2-1 to 2-7). The FSS will include the entire FNR facility, such asthe building, systems, and any other areas, as necessary. Residual radioactivity present inthese structures and components will be decontaminated and/or decommissioned to levels thatwill allow for the unrestricted use of this site.

Figure 2-1 UM FNR Site Plan Figure 2-2 FNR Basement Figure 2-3 FNR First Floor Figure 2-4 FNR Second Floor Figure 2-5 FNR Third Floor Figure 2-6 FNR Fourth Floor Figure 2-7 East-West Cross Section of Reactor Pool 3.0 EVALUATIONThe NRC staff has reviewed the licensee's proposed actions to decontaminate, dismantle, anddispose of component parts of the FNR, and to perform an FSS. In addition, the staff's review focused on the licensee meeting the regulatory requirements discussed in Section 2.1 aboveand included consideration of the following:*management responsibilities/commitments and personnel qualifications to continuefollowing applicable regulations, regulatory guides, standards, and health and safety plans, including procedures

  • use of appropriate equipment and instrumentation, radiation survey methods, training,personnel dosimetry, and radioactive waste disposal
  • the plan to develop and perform the FSS of the facility
  • the commitments needed to implement an adequate quality assurance plan
  • the methods that the licensee will use to meet the radiological release criteria3.1 Decommissioning Alternative The licensee's stated objective of decommissioning the FNR is the release of the site forunrestricted use. As such, the licensee selected DECON as the preferred decommissioning alternative needed to accomplish the stated objective. The licensee will decontaminate facility equipment and structural components to minimizeradioactive waste. Structural portions of the building and materials found to be radiologically contaminated and/or activated will be decontaminated, sectioned and removed, and/orprocessed, as necessary. These activities will be followed by an extensive and comprehensiveFSS to demonstrate compliance with cleanup criteria, and thus allow for release of the site for unrestricted use. To support license termination, the licensee will document the results of thisFSS in a report to be submitted to the NRC for review and approval.3.1.1 Conclusions The NRC staff has concluded that the choice of DECON and associated proposed plans meetthe provisions of 10 CFR 50.82(b)(4)(i) for decommissioning without significant delay and are, therefore, acceptable.3.2Controls and Limits on Procedures and Equipment to Protect Occupational and PublicHealth and Safety3.2.1 Project Management Structure 3.2.1.1 Decommissioning Organization and ResponsibilitiesThe licensee will continue to retain ultimate responsibility for full compliance with the existing NRC reactor license and the applicable regulatory requirements during decommissioning. The responsibility for the decommissioning is assigned to the Executive Vice President andChief Financial Officer. The Executive Vice President has established, through the AssociateVice President for Facilities and Operation, a project organization to oversee thedecommissioning of the FNR as shown in Figure 3-1. The Director of Occupational Safety and Environmental Health leads the FNR project staff and is responsible for the facility's license andauthorizing the expenditure of funds on decommissioning activities. The reactor manager remains responsible for ensuring that decommissioning-related activities are conducted in a safe manner within the limitations of the facility's license and in compliance with applicableFederal, State, and local regulations. The radiation safety officer (RSO), who is organizationally independent of the reactor manager, remains responsible for radiological safety at the facility. A safety review committee, chaired by a representative of the Vice President for Research, is responsible for overseeing decommissioning activities to ensure they are performed safely and in accordance with all applicable license requirements and Federal, State, and local regulations. Figure 3-1 Organization Chart for the FNR Decommissioning ProjectPresidentRegentsUniversity of Michigan Executive Vice President Chief Financial Officer Vice President Research Chair, Review Committee AssociateVice PresidentFacilities & Operations DirectorOccupational Safety and Environmental Health Reactor ManagerReview Committee Radiation Safety Officer ReactorStaff Safety Staff Prime Contractor Project Manager Health Physics Supervisor Prime ContractorStaff Subcontractors, Testing Laboratories, Vendors, Shippers, etc. Techn ical, Operational, Quality & Licensing Management Safety & Environmental Management 3.2.1.2 Key Licensee PositionsThe licensee will maintain the key management positions described below to support thedecommissioning of the FNR.The Director of Occupational Safety and Environmental Health (Director) has oversightauthority and is responsible for the following:*the facility's license (compliance and amendments)*successful completion of decommissioning activities
  • authorizing the expenditure of funds for decommissioning
  • requesting termination of the license for the FNR
  • approval of contractors, subcontractors, and consultants
  • approval of budgets and schedules
  • serving as technical spokesman for UM on decommissioning activities
  • ensuring that the conduct of decommissioning complies with all applicable licenses andregistrations held by UM and with compliance to applicable Federal, State, and local regulatory requirementsSection 5.2 of the DP lists proposed changes to the TSs to update the qualifications for thisposition.The reactor manager has responsibility for the following:*controlling and maintaining safety and protection of the environment duringdecommissioning*determining facility staffing and organization*ensuring that decommissioning activities are within budgetary and schedule requirements
  • reporting performance to the Director and the safety review committee
  • approving changes to the facility that satisfy the equivalent requirements of 10 CFR 50.59,"Changes, Tests and Experiments," contained in the license*providing licensing interface with the NRC, Michigan Department of Environmental Quality,and other regulatory agencies*providing technical oversight and guidance
  • reviewing work procedures, radiation work permits (RWPs), and job hazard analyses (JHAs) *ensuring that shipments of radioactive/hazardous materials are prepared and transportedsafely and in accordance with all applicable regulations and requirements of the receiver*acting as interface between contractor, subcontractors, or consultants and the Director orsafety review committee*coordinating staff, contractor, subcontractor, or consultant activities
  • providing technical support to the Director and safety review committee
  • ensuring that all staff, contractors, and other UM staff supporting decommissioningeffectively implement all quality assurance program(s) requirements*investigating off-normal occurrences or audit findings, scheduling corrective actions,including measures to prevent recurrence of significant conditions adverse to quality, and notifying the Director and each safety review committee member of action taken or planned

to be taken*Assisting the Director in ensuring that decommissioning activities comply with all applicablelicense requirements and with applicable Federal, State, and local regulationsSection 5.2 of the DP lists proposed changes to the TSs to update the qualifications for thisposition.The RSO is responsible for the following:

  • maintaining the radiation safety and health aspects of programs or procedures and ensuringcompliance with programs or procedures*determining facility radiation safety staffing and organization*reviewing work procedures, RWPs, and JHAs in situations that could affect potentialradiation exposure or safety*providing technical support to the Director and safety review committee
  • ensuring procedures and practices are established to ensure that radiation exposures tothe public and facility personnel are kept at as low as reasonably achievable (ALARA) levels*identifying locations, operations, or conditions that have the potential for significantexposures to radiation or radioactive materials and initiating actions to minimize or eliminate unnecessary exposures*monitoring contractor and subcontractor health physics coverage of decommissioningactivities*monitoring collective dose for decommissioning activities *ensuring the implementation of industrial safety, industrial hygiene, and environmentalprotection programs that comply with all applicable license requirements and with applicable Federal, State, and local regulations Section 5.2 of the DP lists proposed changes to the TSs to update the qualifications for thisposition.3.2.1.3 Decommissioning Prime Contractor The licensee provided its criteria for the selection of a prime contractor to manage andsupervise all or part of the FNR decommissioning project. The selected prime contractor willmanage and supervise operations and services such as characterization, dismantlement, decontamination, waste handling, and quality assurance. UM will select the prime contractorthrough an evaluation of the following criteria:*the prime contractor's ability to perform the required task as demonstrated by the quality ofinformation provided in a statement of qualification package*qualifications of key individuals, including but not limited to the key contractor individualsidentified in this section, based upon internal and license requirements*past performance of the contractor and identified key subcontractors with respect tocompliance with all Federal, State, and local regulations*safety record of the contractor and key subcontractors
  • relevant experience of contractor and key subcontractors, particularly with decommissioningof research reactors*references from owners and Federal, State, and local authorities on previousdecommissioning projects for which the contractor and key subcontractors participated*example work products (e.g., RWPs, JHAs, characterization studies, work packages, qualityassurance procedures, etc.) provided by the contractor and key subcontractors*financial qualifications of the contractor and key subcontractors to complete the project The prime contractor will establish and maintain a project manager who will serve as the overallproject manager and a vital member of the project team. The prime contractor will alsoestablish and maintain a health physics supervisor to be responsible for providing basic radiation safety support for contractor and subcontractor activities. The prime contractor may retain subcontractors or hire consultants to help in the performance of all or part of the FNR decommissioning project with the prior approval of the Director.3.2.1.4 Safety Review Committee UM will establish a safety review committee to review decommissioning activities and advise theDirector in matters relating to the health and safety of the project. The safety review committee (as detailed in the DP) will be composed of a chair and a minimumof three members and alternates. The Vice President for Research will appoint the membersand alternates. The safety review committee chair will be appointed from the UM faculty, shallhave a degree in engineering or a scientific field, and will have a thorough understanding of thedecommissioning project. The remaining members of the safety review committee (including alternates) will collectively represent a broad spectrum of expertise appropriate for thedecommissioning of the FNR and may be either from within or outside UM. The safety review committee will meet at least semiannually throughout the duration ofdecommissioning until completion of the FSS. After completion of the FSS, the safety review committee will meet as necessary to review or approve such matters as desired by thecommittee chair, the Director, reactor manager or the RSO. The safety review committee willhave approval, review, and audit functions as described below.The safety review committee will approve the following:*proposed changes in the license or TSs
  • proposed changes to the facility that can be implemented without the prior approval of theNRC in accordance with 10 CFR 50.59*proposed changes in the DP that can be implemented without the prior approval of the NRC*new procedures and proposed changes to the procedures for the following activities whichwill be in effect and followed:-normal operation of all systems structures or components described in the TSs orwhich are important to safety-actions for responding to emergency conditions involving the potential or actualrelease of radioactivity, including provisions for evacuation, reentry, recovery, and medical support-actions to be taken to correct off-normal events and specific malfunctions of systems,structures, or components described in the TSs or which are important to safety-activities performed to satisfy a surveillance requirement contained in the TSs

-radiation and radioactive contamination control

-physical security of the facility

-implementation of the quality controls for the calibration and response testing ofradiation instrumentation used for direct measurement in support of characterization, the FSS, or other quality assurance activitiesThe safety review committee, in its review function, will consider the following:

  • regulatory violations and reportable occurrences made pursuant to license and regulatory requirements*audit reports issued by a member or subcommittee of the safety review committeedeveloped to satisfy any requirement of the committee's audit function*plans for the following decommissioning activities prior to their implementation:-any activity which could compromise the structure and integrity of the reactor pool orthe primary coolant system while pool water is relied upon for shielding of irradiatedreactor components-the dismantlement of the irradiated reactor components in preparation for disposal-the movement of any heavy objects greater than 5 tons in weight

-any activity that could compromise the structural integrity of the post-and-beamstructure that supports the reactor building-any activity that will result in the direct release of radioactivity from the facility to thesanitary sewer or a navigable waterway-the draining of the reactor pool

-the decontamination or dismantlement of the reactor pool structure

-any activity for which it is estimated that the cumulative radiation exposure for theactivity will exceed 1 person-rem, or an individual radiation exposure to either anoccupationally exposed person or a member of the public that could exceed 20 percent of any applicable exposure limits of 10 CFR Part 20-any activity, known or anticipated by the safety review committee, which it requests toreview, subject to the approval of the DirectorThe safety review committee, as an audit function, will ensure that the following areindependently monitored or audited:*decommissioning operations to ensure they are performed safely and in accordance with allapplicable licenses held by UM and in compliance with applicable Federal, State, and local regulatory requirements *the quality assurance program to verify that performance criteria are met as well as todetermine the effectiveness of the program in satisfying the quality assurance requirements of the decommissioning plan and 10 CFR Part 71, "Packaging and Transportation of Radioactive Material"3.2.1.5 Conclusions The licensee has committed to maintaining an adequate organizational structure to oversee andsafely manage the decommissioning of the FNR. The staff has determined that the project management structure for the decommissioning of the FNR is consistent with the guidanceprovided in Appendix 17.1 to NUREG-1537, Revision 0, "Guidelines for Preparing andReviewing Applications for the Licensing of Non-Power Reactors," issued February 1996 (Ref.

6). The management practices described by UM give reasonable assurance that it will continueto be responsible for overall supervision, compliance with regulations, and the health and safety of the public. Therefore, the staff concludes that the proposed project management structure isacceptable.The prime contractor is an integral part of the organization. The licensee intends to choose theprime contractor using the selection criteria presented above. The staff has reviewed these criteria, which cover all skill areas necessary for successful decommissioning project management and performance. Therefore, the staff concludes there is reasonable assurance that the licensee will select a prime contractor with adequate qualifications to support safedecommissioning of the FNR.The staff reviewed and compared the licensee's organizational and control structures. Basedon this review, the staff concludes that the licensee has in place an acceptable organizational structure to safely control the decontamination and dismantlement of the FNR.3.2.2 Occupational and Public Health and Safety 3.2.2.1 Radiation Protection 3.2.2.1.1 ALARA Program The licensee committed to control decommissioning in accordance with the enhancedrequirements of a health physics program and incorporate provisions for reducing individual and collective radiological exposures to ALARA levels. The RSO is responsible for ensuring the establishment of procedures and practices. The FSS plan will fully discuss the implementationof the ALARA requirement of 10 CFR 20.1402 (discussed in Sections 2.1.5 and 4.0 of the DP)to ensure that the ALARA principle is applied to radiation exposures to the public and facilitypersonnel.3.2.2.1.2 Methods for Occupational Exposure ReductionThe licensee presented various methods that will be implemented during the decommissioningproject work to ensure that occupational exposure to radioactive materials is minimized. The methods include the use of RWPs, special equipment, techniques, and other practices as described in the DP. RWPs for jobs with low dose commitments will require approval at thehealth physics technician or health physics supervisory level. The RSO must approve RWPs for jobs with potentially high dose commitment or significant radiological hazards.The health physics organization will ensure that radiation, surface radioactivity, and airbornesurveys are performed as required to define and document the radiological conditions for each job. The licensee committed to instituting process or other engineering controls, as the preferred methods, to maintain exposures to radiation and radioactive materials to ALARA levels. These processes/engineering controls include use of the following:*shielding to reduce the intensity of external sources of radiation *containment and confinement structures to prevent/reduce the potential for generatingairborne radioactivity *ventilation systems to remove airborne radioactivity from the work environment In addition to these ALARA measures, Section 3.1.1.1 of the DP discusses other controls thatwill be employed during decommissioning of the FNR. 3.2.2.1.3 Control and Storage of Radioactive Materials The licensee will continue to rely on the existing health physics program to minimizeoccupational radiation doses during decommissioning operations in a manner that achieves the following:*deters the inadvertent release of radioactive materials to uncontrolled areas*ensures that personnel are not inadvertently exposed to licensed radioactive materials

  • minimizes the volume of radioactive waste generated during licensed activities3.2.2.1.4 Conclusions The licensee has had extensive experience in radiation protection that is directly applicable todecommissioning while operating the reactor facility. The prime contractor will provide furtherexperience and resources under the direction of UM. Based on the review of the DP, the staff concludes that the licensee's ALARA program is acceptable.3.2.2.2 Health Physics Program The DP notes that the existing health physics program at the FNR will remain under the controland authority of UM. Furthermore, the licensee will revise the health physics program asnecessary to ensure that it will continue to satisfy the following radiation protection programcommitments during decommissioning:*minimize the radiological impacts to workers, the public, and the environment
  • monitor radiation level and radioactive materials
  • control distribution and releases of radioactive materials
  • maintain potential exposures to the public and occupational radiation exposure to individualswithin the limits of 10 CFR Part 20 and at ALARA levels3.2.2.2.1 Radiation Exposure UM management committed to minimize exposure of individuals to radiation or radioactivematerials to ALARA levels. To support this commitment, the licensee will subject individualsconducting decommissioning activities to administrative controls for radiation exposure, which will be based on the requirements contained in 10 CFR Part 20 and may be used to ensurecompliance with the annual dose limits and for maintaining exposures at ALARA levels. The following administrative limits apply to FNR decommissioning activities:*employees and contractors-total effective dose equivalent (TEDE) less than or equal to 2.0 rem/year

-total organ dose equivalent less than or equal to 2.0 rem/year

-lens of the eye dose equivalent less than or equal to 2.0 rem/year

-shallow dose equivalent less than or equal to 2.0 rem/year*embryo/fetus (declared pregnant worker exposure)-TEDE less than 0.1 rem over the duration of the pregnancy*visitor, member of the UM community, and member of the public-TEDE less than 0.05 rem/yearThe licensee has written procedures that define administrative limits that are established atlevels less that the allowable occupational dose limits specified in 10 CFR 20.1201, "Occupational Dose Limits for Adults." Furthermore, prior authorization to exceed these administrative limits for any radiation worker will be obtained, in writing, from the licensee'sRSO.The licensee will perform personnel monitoring of occupational radiation exposure from externalsources through the use of individual monitoring devices as required by 10 CFR 20.1502, "Conditions Requiring Individual Monitoring of External and Internal Occupational Dose." The licensee commits to, at a minimum, on an annual basis, or whenever changes in worker exposures warrant, performing an external exposure evaluation to ensure that personnel monitoring of occupational radiation exposure from external sources is in compliance with 10 CFR 20.1502(a). Dosimeters that require processing (e.g., thermoluminescent or optically stimulated luminescence dosimeters) will be provided by UM and will be processed by adosimetry processor accredited by the National Voluntary Laboratory Accreditation Program.The licensee will determine occupational internal exposure from licensed radioactive materialsto an individual through monitoring of the quantities of licensed materials in the air collected through air samples, in vitro or in vivo bioassay techniques, or a combination of air monitoring and bioassay as allowed by 10 CFR 20.1204, "Determination of Internal Exposure," and required by 10 CFR 20.1502 (b). If respiratory protection equipment is used for protection against airborne radioactive material, then the licensee will evaluate the actual intakes, takinginto account the protection factors assigned to the type of respiratory protection employed as allowed by 10 CFR 20.1204. To ensure compliance with 10 CFR 20.1502(b), bioassay for intakes of licensed materials may be performed for licensee personnel with the greatest potential for intake at a sample frequency appropriate for the pulmonary retention class (days, weeks, years).When exiting restricted areas that have known removable contamination or the potential for removable contamination, site personnel will monitor their hands and feet for contamination inaccordance with internal procedures. If contamination is detected, then the site personnel willcheck the exposed areas of the body and clothing. Site personnel leaving potentiallycontaminated areas will periodically monitor their hands and feet for contamination, consistentwith the nature and quantity of the radioactive materials present.The licensee will continue to measure the concentrations of radioactive material released fromthe facility in gaseous effluents. The dilution factor of 400, taken from previous safety analysessubmitted to the NRC and contained in the TSs, continues to apply to the FNR exhaust and thePML stack exhausts. UM may also use other options for showing compliance with the annual dose limit to an individual member of the public from concentrations of radioactive material released from the facility in gaseous effluents, as allowed by 10 CFR 20.1302, "Compliancewith Dose Limits for Individual Members of the Public."To ensure compliance with the requirements of 10 CFR Part 20, the licensee will conti nue tomeasure the concentrations of radioactive material released from its facility in liquid effluents. UM may also use other options for showing compliance with the annual dose limit to an individual member of the public from concentrations of radioactive material released from the facility in liquid effluents as allowed by 10 CFR 20.1302.3.2.2.2.2 Surveys and Monitoring The licensee will perform radiation surveys and monitoring in accordance with the existingradiation protection program and as necessary to support work activities in areas with the potential for exposure to radiation or radioactive materials. The licensee will assess the effectiveness of controls to minimize or eliminate radiation exposures in the following two ways:(1)direct measurement of the external radiation or the radioactive material intake an individualreceives(2)measurement of the radiological conditions in the area(s) occupied by the individual Levels and extent of direct radiation and radioactive materials in any work area will bemeasured and assessed in accordance with the licensee's health physics program. These measurements will include, as a minimum, the following:I.direct dose rate measurementsII.surface contamination measurements (fixed and removable)

III.airborne radioactive material measurementsThe licensee will ensure that instruments and equipment used for these measurements arecalibrated for the radiation type to be measured on frequencies as listed in Section 3.1.2.4 ofthe DP.3.2.2.2.3 Exposure Control The licensee defines restricted areas based on the known or suspected hazard potential fromradiation sources that have been defined from measurement or inferred from process knowledge. Radiation exposures to an individual entering such an area may be assessed from any combination of the following:IV.direct radiationV.surface contamination (fixed and removable)

VI.airborne contamination3.2.2.2.4 Control of Exposure to Direct Radiation Control of exposure to individuals from direct radiation is based on two elements, as defined inthe DP:(1)measurement and assessment of the location and strength of the radiation sources(2)control of the individual's access to those radiation sourcesRoutine monitoring of the levels and extent of radiation and radioactive materials is a key part ofthe licensee's health physics program. The program also involves measuring and assessing the levels and extent of direct radiation and radioactive materials in work areas. These measurements include direct dose rate, surface contamination, and airborne radioactive material measurements.Before defining control requirements for limiting direct radiation exposure to individuals, thelicensee will determine the location of the radiation sources and the magnitude of the radiation. Direct radiation exposure measurements will be made at the time of decommissioning,concentrating on areas identified as having a worker exposure potential. This survey work also will include specific areas or systems identified by the licensee during work planning beforeproject startup.Based on this measurement and data assessment, the licensee will establish shielding, orbarriers that restrict access to sources of radiation. The licensee will install postings at access points through those barriers based on the potential exposures that an individual could receiveupon entry through the access points or along external surfaces of the barrier, in accordance with regulatory requirements.3.2.2.2.5 Control of Exposure to Surface ContaminationThe licensee will control exposure to individuals from surfaces contaminated with radioactivematerial either by prior decontamination or by using protective equipment for personnel to minimize or limit exposure to the surface material.Prior decontamination for planned work activities is the licensee's preferred method ofcontamination control. However, the licensee will evaluate this practice for ALARA considerations to ensure that exposures resulting from the decontamination/removal do notoffset exposure savings for the planned work activities.The licensee may need to establish controlled surface contamination areas because thecontaminants present at the FNR are primarily beta-gamma emitting activation and fission products. The licensee will use administrative control postings for "contamination areas" and"high contamination areas" as follows: VII.contamination area-an area where surface contamination levels exceed therequirements for unrestricted release of a surface, but are less than 100 times the surface values in Table 3-1 of the DPVIII.high contamination area-an area where surface contamination levels exc eed 100times the surface values in Table 3-1 of the DPWhen decontamination is impractical or ineffective, personal protective equipment (PPE) will beused to protect individuals from potential radiation exposures attributable to surface contamination. The licensee will consider radiological conditions, type of work to be performed,potentially stressful environmental conditions, physical condition of surfaces, and duration of the activity in determining the appropriate PPE. If the potential for exposure to airborne radioactivity at levels in excess of 12 derived airconcentration-hours in a workweek is encountered, the licensee will require workers to don full-face respirators for work activities that will be performed in these areas.As appropriate, the licensee will employ contamination control measures that include, but arenot limited to, the following:IX.local containment barriers such as designed barriers, glove bags, containers, and plasticbags to prevent the spread of radioactive materialX.physical barriers such as Herculite sheeting, strippable paint, tacky-mat step-off pads,absorbent pads, and drip funnels to limit contamination spread3.2.2.2.6 Control of Exposure to Airborne ContaminationIf air monitoring results indicate levels of airborne radioactive materials in excess of NRC-prescribed levels, the licensee will post the area as an "airborne radioactivity area" at accesspoints, per the definition in 10 CFR 20.1003, "Definitions."When it is not practical to employ the engineering controls described previously, or when thesecontrols are not sufficient to maintain the airborne radioactivity levels below those defining an "airborne radioactivity area," the licensee will then require the use of respiratory protectionequipment for individuals entering this work environment. When respiratory protection is required, it will be as described in a respiratory protection program satisfying the requirementsof 10 CFR Part 20, Subpart H, "Respiratory Protection and Controls to Restrict Internal Exposure in Restricted Areas." The licensee's program will include worker training and medicalqualification requirements for use and descriptions of the following:XI.respiratory protection equipment to be usedXII.air monitoring requirements to support the use XIII.bioassay program to evaluate the effectiveness of useXIV.equipment cleaning, testing, and maintenance requirements3.2.2.2.7 Radiation Monitoring Equipment The licensee will maintain a sufficient inventory and variety of instrumentation onsite to facilitate effective measurement of radiological conditions and control of worker exposure consistent withALARA principles and to evaluate the suitability of materials for release for unrestricted use. The licensee will employ radiation monitoring equipment capable of measuring the range ofdose rates and radioactivity concentrations expected to be encountered during remediation and decontamination activities to the minimum values required for release of materials for unrestricted release. The licensee committed to calibrate radiation monitoring equipment at the intervals prescribedby the manufacturer-annually, or prior to use as discussed in Table 3-2 of the DP. Radiation monitoring equipment will be calibrated using standards traceable to the National Institute ofStandards and Technology (NIST). Calibration information will be clearly marked on theinstrument. In addition, survey instruments and equipment will be operationally tested dailywhen in use. 3.2.2.2.8 Conclusions The licensee has a mature health physics program capable of protecting workers andminimizing the levels of radiation exposures that may be encountered during decommissioning of the FNR. As such, the NRC staff finds that there is reasonable assurance that theimplementation of the procedures and guidance of the health physics and ALARA programs willminimize the radiation exposure of workers and the public. The staff concludes that the licensee's health physics program is acceptable and meets the requirements in 10 CFR 20.1101, "Radiation Protection Programs."Based on the review of the respiratory protection program proposed in the DP, the staffconcludes that the licensee has the necessary organizational structure and management controls to establish and maintain a program that meets the requirements of 10 CFR Part 20,Subpart H.3.2.2.3 Control of Radioactive Materials The licensee will survey all materials leaving a restricted area to ensure that such equipment,materials, and items do not contain detectable quantities of radioactivity. The licensee's surveys will incorporate the guidance in NRC Circular No. 81-07, "Control of RadioactivelyContaminated Material," dated May 14, 1961 (Ref. 7), and Information Notice No. 85-92, "Surveys of Wastes Before Disposal from Nuclear Reactor Facilities," dated December 2, 1985(Ref. 8).For items that may be contaminated with beta-gamma emitting activation and fission products,the licensee will use the following survey methods:XV.materials and equipment-direct frisking with a portable Geiger-Mueller detector (e.g.,Ludlum Model 44-9, Eberline Model HP-210, or equivalent) having a minimum level of detection above background of less than or equal to 5000 disintegrations per minute (dpm) per 100 square centimeter (cm 2)XVI.smear samples-analysis with a Geiger-Mueller detector (e.g., Ludlum Model 44-9,Eberline Model HP-210, or equivalent) having a minimum detection level above background of less than or equal to 1000 dpm per 100 cm 2 XVII.bulk materials (e.g., sand and soil)-analysis of representative sample(s) using a high-resolution gamma spectroscopy system having a lower limit of detection abovebackground of less than or equal to 0.18 picocurie (pCi) per gram for cesium (Cs)-137 XVIII.background-equivalent gamma activity-an unshielded gamma ray dose measured1 meter from any surface, not to exceed 5 microrem per hour above backgroundThe licensee may develop additional methods for release of surface contaminated materials,which would be subject to the minimum detection levels in Table 3-3 of the DP. Detection sensitivities of instruments and techniques may be determined using the guidance contained in the NUREG-1575, Revision 1, "Multi-Agency Radiation Survey and Site Investigation Manual(MARSSIM)," issued August 2000 (Ref. 9), and NUREG/CR-1507, "Minimum DetectableConcentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions," issued June 1997 (Ref. 10). The licensee may relocate equipment and materials to areas of lower ambient background for the conduct of release surveys. The licensee willrelease materials only if a survey using a method identified above does not identify any discernable radioactivity above background from licensed materials.In evaluating equipment and materials for fixed or smearable licensed radioactive materials, thelicensee will not release items painted with other than the original manufacturer's paint unlessclear process knowledge demonstrates that the paint was applied to a clean surface containingno discernable radioactivity from licensed materials prior to its use in a restricted area.

Following the satisfactory completion of a survey satisfying the requirements listed above, the licensee's RSO may approve the release of items for which it cannot be demonstrated that paint was applied to a clean surface.If the potential exists for contamination on inaccessible surfaces, the licensee will assume theequipment to be internally contaminated unless (1) the equipment is dismantled allowingaccess for surveys, (2) appropriate tool or pipe monitors satisfying the survey requirements listed above are used to provide confidence that no licensed radioactive materials are present, or (3) it may readily be concluded that surveys from accessible areas are representative of the inaccessible surfaces (i.e., surveying the internal surface of both ends of a pipe from a nonradioactive process system with cotton swabs would be representative of the inaccessibleareas).Personal effects (e.g., notebooks, pens, flashlights) that are hand-carried into a restricted areaare subject to the same survey requirements as the individual possessing the item. The licensee may transfer licensed radioactive materials to other locations within the control ofUM as allowed by appropriate radioactive material licenses issued by the NRC. The licenseemay transfer licensed radioactive materials to other locations outside UM that possess the appropriate radioactive material licenses issued by the NRC, an Agreement State, or areotherwise authorized to possess such radioactive material (e.g., DOE sites, foreign research reactors).The licensee will revise its existing health physics program (described in Section 3.2.2.2)revised as needed in accordance with the internal approval and change control provisions discussed in Section 2.4 of the DP.3.2.2.3.1 Conclusions The NRC staff finds that the licensee has an adequate program for assessing whetherequipment, tools, materials, and items contain detectable quantities of radioactivity prior to their unrestricted release. The NRC staff bases its conclusion on the licensee's commitment to adhere to NRC Circular No. 81-07 for making this assessment. Furthermore, based on commitments from the licensee, volumetric releases of bulk materialsfor unrestricted release will be limited solely to soil and sediment, but will not include rubblizedconcrete or other similar manmade items. Should the licensee elect to seek approval to release such manmade items in the future, the NRC staff noted that an exemption under10 CFR 20.2002, "Method for Obtaining Approval of Proposed Disposal Procedures," would be required based on a site-specific radiological dose assessment reflective of the site where such items would ultimately be dispositioned. Based on information and commitments provided by the licensee, the NRC staff concludes thatthe licensee has an adequate health physics program for ensuring that equipment, tools, materials, and items do not contain detectable quantities of radioactivity prior to their unrestricted release in support of the decommissioning of the FNR.3.2.2.4 Dose Estimates The licensee estimated the total occupational exposure to complete the FNR decommissioningproject to be 4.8 person-rem. The licensee based this dose estimate on characterization data and professional judgment that took into account the individual work activity durations and workcrew sizes estimated by the UM contractor.While the dose estimate provided in the DP is for planning purposes only, the licensee willdevelop detailed exposure estimates and exposure controls in accordance with the requirements of the ALARA program. The licensee will perform the actual estimate of exposurethat may be incurred in the course of decommissioning during detailed planning of the decommissioning activities. The licensee estimated that the dose estimate to members of the public as a result ofdecommissioning activities will be negligible. The licensee based this conclusion on the factthat the area immediately surrounding the facility is under UM control and because the areawhere decommissioning activities are taking place is fully contained within the facility (with theexception of loading and unloading shipments of equipment and radioactive materials). This is consistent with the negligible (less that 0.1 person-rem) dose estimate provided for the reference research reactor in NUREG-0586, "Final Generic Environmental Impact Statement onthe Decommissioning of Nuclear Facilities," issued 1988 (Ref. 11).3.2.2.4.1 Conclusions Based on the NRC staff's review of information provided by the licensee, the licensee'sestimates for occupational and public dose during decommissioning activities are reasonable and well below the radiation dose levels permissible to workers and members of the public as specified in 10 CFR Part 20. Therefore, the NRC staff concludes that the FNR can be safelydecommissioned to levels allowed under such Federal regulations.3.2.2.5 Radioactive Waste Management Section 3.5.3 of this document discusses quality assurance provisions for the transport oflicensed radioactive material.3.2.2.5.1 Fuel Removal To support issuance of a possession-only license, the NRC approved a revised TS thatprohibits the licensee from operating the FNR and possessing reactor fuel. The agency based this licensing action upon the licensee's decision to remove all irradiated and unirradiated reactor fuel from the site. All irradiated reactor fuel was removed from the FNR and returned to the DOE's Savannah River Site between October and December 2003. In addition, the licensee returned all unirradiated reactor fuel to DOE through BWXT Technologies in August

2003.3.2.2.5.2 Radioactive Waste Processing The licensee stated that in general, system components will not be decontaminated onsite. Slightly contaminated items may be decontaminated onsite if it is determined that a componentor portion of a component can be safely and economically decontaminated. Onsite personnel, including staff, contractors, or specialty contractors will decontaminate such items using techniques and materials within the capabilities of those personnel as determined by UM. Experienced offsite vendor(s) may also be used to decontaminate the components if they can be safely and economically decontaminated as determined by UM. Currently, the licensee intends dismantle the contaminated piping systems and dispose of the material as radioactivewaste or to decontaminate and free release those materials.The licensee's decommissioning of the FNR will result in the generation of solid and liquid low-level radioactive waste, mixed waste, and hazardous waste. Solid radioactive wastes include neutron-activated materials, contaminated materials remaining in the reactor building, and those items necessarily contaminated onsite during the remediation activities. The licensee anticipates limited, if any, soil remediation that would result in generating solid radioactivewastes. Liquid low-level radioactive waste includes the water in the reactor pool and the associated piping as well as contaminated water generated during remediation activities. The licensee stated that no gaseous radioactive waste exists because the reactor has been shut down for more than 9 months and all radioactive gases have since decayed.The licensee will perform handling, staging, and shipping of packaged radioactive waste inaccordance with all applicable Federal, State, and local regulatory requirements. The majority of the solid waste expected to be generated during decommissioning of the FNR isexpected to consist of Class A low-level radioactive waste. Onsite radioactive waste processing will include waste minimization, volume reduction, segregation, characterization, neutralization,stabilization, solidification, and packaging. Wastes may be shipped to a licensed processingfacility for survey and release or decontamination and release, or may be disposed of directly ata licensed facility in accordance with its radioactive material license. A manifest consistent with the proper waste classification will accompany each shipment of radioactive waste, as specifiedin Section I, "Manifest," of Appendix G, "Requirements for Transfers of Low-Level Radioactive Waste Intended for Disposal at Licensed Land Disposal Facilities and Manifests," to10 CFR Part 20.3.2.2.5.3 Radioactive Waste Disposal Low-Level Liquid Radioactive Waste The licensee plans to dispose of approximately 50,000 gallons of low-level radioactivelycontaminated water currently contained in the reactor pool and associated piping by treatmentand discharge to the public sewer system. The City of Ann Arbor operates the sewer system inaccordance with Federal, State, and local regulatory requirements. Any additional radioactively contaminated water generated during remediation activities may also be similarly discharged in accordance with all applicable regulatory requirements. The licensee will monitor and process the low-level radioactively contaminated water from thereactor pool and associated piping using techniques consistent with the licensee's health physics and ALARA programs. The liquid waste generated during licensee remediation activities will be monitored andprocessed prior to discharge. During demolition activities, installed plant equipment used to process liquid radioactive waste may be removed. Therefore, temporary filtration units or demineralizers may be used as the primary means of treatment. Any temporary liquid treatment system necessary to ensure that disposal requirements are met will be connected totanks for storage of processed water prior to discharge. Once the licensee verifies that the stored processed water meets the allowable discharge limits specified in the TSs, the water may be subsequently released. The existing effluent monitoring instrumentation will be used tomonitor discharges of liquid effluent as required and to demonstrate compliance with the TSs and other applicable regulations.Makeup water used for flushing will typically originate from the existing potable water supply. The effluent stream(s) from such activities will be processed as above, by filtration anddemineralization.The licensee will implement measures to treat liquid radioactive waste to maintain workerradiation doses within the required regulatory limits during decommissioning of the FNR. As such, filters will be replaced as appropriate and system components will be positi oned orshielded. If radioactively contaminated water is discharged to the sanitary sewer, the discharge piping willbe resurveyed and remediated as necessary.The DP details the following disposal options:

XIX.Low-level radioactively contaminated water may be evaporated onsite. For suchactivities, the facility's effluent monitori ng system is equipped to monitor airborneradioactive effluents in accordance with the TSs and applicable regulations. XX.The licensee may choose to use a licensed radioactive waste processor(s) to providespecialized services for reducing the volume of or treating radioactive liquid waste.

Such services may include demineralization, direct incineration, ground application, evaporation, and survey and release. UM may also elect to transfer all or some of the liquid radioactive waste from decommissioning to a licensed waste processor.XXI.Currently, the licensee does not plan to use chelating agents in any chemicaldecontamination activities for FNR systems or structures. Radioactive wastescontaining chelating agents will be generated only if necessary and will be minimizedto the fullest extent possible.XXII.The licensee will return tritium-loaded heavy water, owned by DOE, to the DOESavannah River Site for processing and reuse.Solid Radioactive Waste While the majority of solid waste generated during the decontamination and dismantlement ofactivated and contaminated systems, structures, or components is expected to consist ofClass A low-level radioactive waste, the licensee noted that information on the estimated curie(Ci) content and waste volume for this decommissioning project is extremely limited at this time.

Additional information is therefore required to determine the specific waste classification. The estimates of waste volumes are conservative and do not account for any volume reduction techniques. In addition, the estimates assume only direct burial rather than allowing for decontamination and possible free release. The licensee is planning a number of measures to reduce the volume of solid radioactive wastethat will require disposal at a licensed burial facility. The primary components of the solid wasteto be generated by the decommissioning of the FNR facility are expected to be disposed assummarized below and in Section 3.2.4 of the DP: I.Irradiated reactor hardware may require size reduction to facilitate loading. Irradiatedreactor hardware will be loaded into a high-integrity container (HIC) or liner, then placed inan approved, shielded shipping cask for transport and subsequent direct burial at the licensed land disposal facility in Barnwell, South Carolina. The current estimate for thevolume of irradiated reactor hardware requiring burial at Barnwell is 300 cubic feet.

II.The contami nated systems piping and equipment will be segmented. As cuts are m ade, asuitable cover will be placed on open ends to preclude the spread of contamination. Material that can be economically dismantled and decontaminated will be appropriatelyhandled onsite or sent to a vendor facility for decontamination. Material that c annot beeconomically decontaminated will be placed in proper disposal containers (e.g., low specificactivity (LSA) containers) and sent to an appropriate processor or burial facility. Thelicensee expects approximately 5300 cubic feet of activated or contaminated material to be generated for processing or disposal.III.Activated or contaminated concrete removed in large sections will be packaged as LSAmaterial in approved shipping containers for direct shipment to the licensed land disposal facility operated by Envirocare of Utah, Inc. An estimated 5200 cubic feet of activated orcontaminated concrete, two-thirds of the concrete comprising the reactor pool, will require disposal in this manner.IV.Dry Active Waste (DAW) consisting of contaminated paper, plastic, coveralls, and similaritems will be packaged as LSA material in approved shipping containers. The licensee willship uncompacted DAW to an offsite vendor for volume reduction and processing if supported by ALARA and cost considerations. When feasible, DAW will be used to fill voidspace in other radioactive waste shipping containers; otherwise, it may be shipped for direct burial. An estimated 300 cubic feet of DAW will require transfer to a licensed wastedisposal facility for postprocessing and disposal.V.Engineering controls such as high-efficiency particulate air (HEPA)-filtered ventilation will berequired to capture potential airborne contaminants. Spent HEPA filters will be c hanged outand treated as DAW. An estimated 25 cubic feet of contaminated filter media will requiretransfer to a licensed waste disposal facility for postprocessing and disposal. VI.Radioactive waste treatment systems will be required to process the liquid waste streamresulting from various decommissioning activities as described above. The licensee will usefiltration and ion exchange processing to remove residual radioactivity in the water. A vendor or the FNR may supply temporary demineralization and filtration systems. Thelicensee estimates the volume of spent resins and filters required to process the water to be less than 400 cubic feet. These resins will be transported to a licensed facility for disposal.VII.The licensee does not expect radioactively contaminated asbestos waste to bepresent, but it may be identified by decommissioning or preparatory activities.

Asbestos material should be transferred to an offsite, licensed radioactive waste processor for compaction or for survey and release. Large items containing asbestos waste may require size reduction before transfer to the offsite, licensed radioactive waste processor.VIII.The only known mixed waste at the FNR is from lead shielding, possibly lead paint,and cadmium. The FNR has approximately 13,000 pounds of contaminated lead, 1,600 pounds of activated lead, 400 pounds of contaminated cadmium, and 20 pounds of activated cadmium. A vendor will encapsulate or otherwise treat thesematerials for ultimate disposal or recycle. The objective of UM is to generate no newmixed waste during decommissioning activities. Procedures currently in place for hazardous and radiological waste management are sufficient to provide the assurance that waste will not be generated arbitrarily and that generated wastes will be disposedof properly.3.2.2.5.4 Method of Estimating Types, Amounts, and Radionuclide Concentrations of Radioactive Waste Generated during DecommissioningThe licensee will derive an estimate of total radioactivity present in systems, structures, orcomponents directly from field radiological measurements, supplemented by analytical data or through computational estimates, as follows:IX.sampling volumetric material to establish ratios of radionuclides present in a structure orcomponent X.direct measurement using sodium iodide (NaI), high-purity germanium, or other detectors toanalyze the gamma spectrum being emitted to identify specific isotopes, establish ratios of isotopes, or to fully quantify isotopesXI.direct measurement of dose rates to support computational methodologies for thedetermination of radionuclides presentXII.direct measurement of similar items for extrapolation via computational methods forinaccessible components or structuresEstimates of the radionuclide concentration in irradiated items may be based on the constituentelements of the material in question and by calculating the duration of exposure and the energies of the incident neutrons. The licensee will use radiological surveys to determine theactivity present within internally contaminated piping and on structures.3.2.2.5.5 Conclusions Based on the review of the licensee's program as described in the DP and the licensee'sexperience, the NRC staff concludes that the licensee's proposed radioactive wastemanagement plans for the UM decommissioning project are acceptable and will conform to NRC regulations.3.2.3 Training Program Because decommissioning activities are much different from typical FNR operations, thelicensee committed to conduct special training for the existing FNR operations staff and the decommissioning personnel. Individuals (employees, contractors, and visitors) who require access to the work areas or radiologically restricted area will receive training commensurate with the applicable regulatory requirements (i.e., 10 CFR Part 19, "Notices, Instructions and Reports to Workers: Inspections and Investigations") for the potential hazards to which they may be exposed. Individuals will also receive continued training, as necessary, to ensure thatjob proficiency is maintained.Personnel will be qualified for their assigned duties prior to performing such work or will beunder the direct supervision of a qualified employee. Personnel performing special processes will be qualified according to specific codes and standards and/or in accordance with nationalconsensus documents. Qualification will include proficiency demonstrated by each individualprior to performing work and periodically assessed throughout the duration of the project.

Qualification also will be demonstrated when required by the designated codes or standards.The licensee will maintain training records that include the trainee's name, dates of training,types of training, test results, protective equipment use authorizations, and instructor's name.Care will be taken to ensure that properly qualified instructors conduct all training. As theprimary criterion, persons responsible for presenting training should have knowledge and experience in the process or subject matter. It is desirable that trainers also have the presentation skills or classroom conduct appropriate to the level of the training being presented. For those with limited experience in conducting training, early instruction should be monitored and feedback should be provided. The licensee provided examples of the various types of training programs applicable todecommissioning activities::I.general employee training-general training for emergency response, spill response,alarms, alarm response, communication systems and channels, waste management, andwaste minimization II.radiation safety training:-general radiological training-training for personnel who are required to enterradiological restricted areas, with the exception of visitors and infrequent support personnel, but are not authorized to perform hands-on radiological work-radiological worker training-training for personnel who require unescorted access toradiological restricted areas and who are authorized to perform radiological job functions-core training-may be accomplished under any program that meets basicrequirements-site-specific training-given to all personnel

-refresher training-given annually to all personnel*hazardous waste operations and emergency response-training for personnel engaged inhazardous substance removal or other activities that potentially expose them to hazardoussubstances and health hazards, which satisfies 29 CFR 1910.120, "Hazardous Waste Operations and Emergency Response"*respirator training and fit testing-training, medical qualification, and fit testing for eachperson who wears a tightly fitting respirator that satisfies the requirements of 10 CFR Part 20, Subpart H, and Regulatory Guide 8.15, "Acceptable Programs for Respiratory Protection" (Ref. 12)*Department of Transportation hazardous materials employee training-training as requiredby 49 CFR Part 172, "Hazardous Materials Table, Special Provisions, Hazardous Materials Communications, Emergency Response Information, and Training Requirements,"

Subpart H, "Training," provided to all personnel involved in the loading, unloading, or handling of hazardous materials, preparing hazardous materials for transportation (including packaging and preparation of manifests), or responsible for the transportation of radioactive materials or operation of a vehicle used to transport hazardous materials (49 CFR 171.8, "Definitions and Abbreviations")*security requirements for offerors and transporters of hazardous materials-training for inthe facility's security plan that satisfies the requirements of 49 CFR Part 172 for allpersonnel involved in the offering of placarded quantities shipments of hazardous materials*hazard communication training-training covering, at minimum, the proper use of materials,the required PPE, and the emergency procedures associated with these materials for all personnel on the hazardous chemicals in their work area, as required by29 CFR 1910.1200(h), including update training whenever a new physical or health hazard is introduced into their work area*hearing conservation training-training on the effects of noise on hearing and the purposes,advantages, disadvantages, and attenuation of various types of hearing protective devices*permit-required confined space entry training-training for personnel if entry into confinedspaces is to be performed*lockout/tagout training-training for hazardous energy control

  • trenching and excavation training-training for the purpose of determining the safety andstability of excavations *fire watch training-training on the proper selection, use, and application of extinguishingagents; characteristics and classification of fires*asbestos abatement training-training on requirements, potential health effects, andcontrols for asbestos abatement*torch/plasma arc cutting, welding, and open flame trainings-training in the use of, andunderstanding the reasons for, protective clothing and equipment, including the need for flame-resistant clothing*tailgate training-routine, short training, given usually at the beginning or end of a regularworkforce briefing, intended to provide a brief review of a safety or programmatic topic, which is applicable to current work activities*other specific mandated training-any other training that may be required by the standardsspecific to the Michigan Occupational Safety and Health Act of 1974 (MIOSHA) or applicable standards before initiating work that may fall within the scope of decommissioning3.2.3.1 Conclusions Based on the review of the licensee's training program as described in the DP, the staffconcludes that the licensee's training program is acceptable. The licensee also recognized thatspecific training would be required to reflect the unique hazards associated with decommissioning operations. While the NRC does not regulate nonradiological hazards asspecified in the Atomic Energy Act, the licensee is aware that personnel involved with decommissioning activities would be subject to training requirements administered by other Federal, State, and local government agencies.3.2.4 General Industrial Safety Program The licensee stated that the RSO, with the cooperation of the full project management team,will be responsible for ensuring that the occupational health and safety requirements for projectpersonnel are met, primarily in terms of compliance with the Occupational Safety and Health Act of 1973 and MIOSHA. Specific responsibilities include establishing training requirementsfor general safe work practices, reviewing plans and procedures to verify adequate coverage of industrial hygiene and safety requirements, conducting periodic inspections of work areas and activities to identify and correct any unsafe conditions and work practices, coordinating industrial hygiene services as required, and advising the Director on industrial hygiene and safety matters and on the results of periodic safety inspections.All personnel working on the FNR decommissioning project will receive health and safetytraining in order to recognize and understand the potential risks to personnel health and safetyassociated with the work at the FNR. The health and safety training also ensures compliance with the applicable regulatory requirements. Personnel will be trained on the plans, procedures,and operation of equipment to conduct work safely on the FNR decommissioning project.The implementation of occupational health and safety requirements for activities involvingpotential hazards that may be encountered during decommissioning will be evaluated thr oughthe use of a JHA. Each JHA will identify all hazards associated with the activity (e.g., fallprotection, hot work, confined space). The licensee will prepare a procedure implementing theJHA that will be subject to the approval requirements discussed in Section 2.4 of the DP. TheJHA allows the project management, project staff, contractor staff, and UM industrial safety personnel (through the RSO or reactor manager) to specify the controls and processes necessary to protect the safety of individual workers, the UM community, and the public. The JHA will act in concert with the RWP, if required, to complete the protection program. Arepresentative of the UM industrial safety staff, the RSO, or the reactor manager will approvethe JHAs. In their absence, the RSO and the reactor manager can delegate this approval authority.3.2.4.1 Conclusions Based on the review of the licensee's proposed industrial safety program as outlined in the DP,the staff concludes the program is acceptable.3.2.5 Radiological Accident Analyses The licensee evaluated radiological accidents that could potentially occur duringdecommissioning of the FNR. This accident analysis considered areas that contain the highestinventories of radioactive material expected to be present during the decommissioning of the FNR. The results of this analysis adequately bounded the radiological impacts that could reasonably occur during decommissioning. As such, a fire, a pool leak, and a tritium-loaded heavy-water spill were the radiological accidents considered to present the highest potentialconsequences.3.2.5.1 Fire The licensee considered the consequences of a fire during decommissioning of the FNR anddid not find them to be significantly different than the consequences of a fire during reactoroperations. The majority of the materials of construction present in the FNR are metals, concrete, or similar noncombustible materials. Upon termination of reactor operation, most of the combustible materials required for reactor operations were removed from the reactor building to further reduce the potential consequences of a fire. The licensee concludes that it is highly unlikely that a fire would start or that a fire could become intense enough to ignite thesetypes of materials (including other combustible materials such as rags, wipes, and anticontamination clothing), and thus result in the release of radioactive material. The licensee stated that dry radioactive waste is normally collected in metal pails with lidslocated throughout the facility. Once full, the dry waste is normally transferred into 55-gallondrums meeting the strong-tight requirement for shipment to a licensed waste processor. Small quantities of dry radioactive waste requiring special handling or segregation are stored in plastic 5-gallon pails. The licensee stated that this practice limits the volume of dry radioactive waste that could be ignited in a fire event to a few pounds and serves to lower the potential for a fire toconsume additional waste collections. The licensee contends that any fire involving dry radioactive waste would be limited to a few microcuries of radioactivity from radionuclides contained in the list of expected radionuclides (refer to Table 2-4 of the DP).During a fire involving dry radioactive waste, the emission of airborne radioactivity from the FNRexhaust stack would continue unless operator action is taken, or upon automatic closure of the ventilation dampers when the radioactivity levels exceed 1 millirem (mrem) per hour at thebuilding exhaust radiation monitor (required by the TSs). The licensee stated that for the purposes of the evaluation, the ventilation dampers were assumed to remain open, and an exhaust stack dilution factor of 400 and an emission rate of a minimum of 8000 cubic feet per minute up the FNR exhaust stack was assumed, for a duration of 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br />.Table 3-1 presents the emissions of individual radionuclides that could be released to theenvironment resulting from a fire without exceeding the airborne effluent concentration (AEC) limits for a full year as specified in Table 2 of Appendix B, "Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage," to 10 CFR Part 20. Table 3-1 Quantities of Individual Expected Radionuclides Producing the Emission ofthe AEC during an 8-Hour FireNuclideIndividual Quantity 1Antimony-125 W class130 mCiBismuth-210m D class0.4 mCi Cadmium-109 W class8.7 mCi Carbon-14 (Monoxide)86 Ci Cesium-134 D class8.7 mCi Cesium-137 D class8.7 mCi Cobalt-60 W class8.6 Ci Europium-152 W, All classes390 mCi Europium-154 W, All classes1.3 mCi Iron-55 W class260 mCi Manganese-54 All classes44 mCi Nickel-59 W class430 mCi Nickel-63 W class173 mCi Scandium-46 Y, All classes13 mCi Silver 108m W class17 mCi Silver 110m W class13 mCi Tritium4.3 Ci Zinc-65 Y, All compounds17 mCi 1Activity = AEC x 28,317 cc/ft 3 x 8,000 cfm x 60 min/h x 8 h x 400 Because these quantities of radionuclides are in the mCi range, the licensee stated that theyare significantly greater than the levels expected in any localized, individual containers of dryradioactive waste (i.e., rags, wipes, and anticontamination clothing). During a fire involving these types of wastes in which the ventilation system is secured shortly after the initiation of thefire, the licensee described that exposure would be limited to those individuals who provide initial emergency fire suppression activities. As such, the bounding consequence of the most credible fire event involving an individual container would result in a maximum exposure of 50 mrem to a member of the public. The construction of the reactor building provides three locations where hot gases from a firewould collect. One of these locations is the area above the reactor pool; the other two locations are just below the third floor on both the east and west sides of the reactor pool. The licenseeconcludes that this inherent design feature aids in the reduction of the concentration ofradioactive materials in the breathing space near the first, second, and third floors of the reactor building, and limits the inhalation of radioactive materials of individuals who provide initial fire suppression activities, those individuals who evacuated the facility, and those individuals whoare required to reenter the reactor building.In the event of a fire, individuals present in the facility may make a reasonable attempt toextinguish the fire using the portable extinguishers provided throughout the facility. If the firecannot be extinguished, the Ann Arbor Fire Department is summoned, as discussed in the FNR Emergency Plan. According to the licensee, it expects only minimal radiological exposures to be incurred by individuals during the short period while attempting to extinguish the fire in the dry radioactive waste or evacuating the area. In addition, fire-fighting personnel responding to a fire potentially involving radioactive materials are trained to use PPE (including adequate respiratory protection, which the licensee concludes would ensure that any internal exposurewould be significantly less that the 50 mrem bounding dose analyzed for a member of the general public).3.2.5.2 Pool Leak In the event of a major leak from the reactor pool, all water loss would be collected by the floordrains or pass through openings in the first floor to the basement of the reactor building. The dimensions of the reactor basement are large enough to allow for the collection of all 50,000 gallons of water from the reactor pool. Loose radioactive contamination in the water's pathway to the reactor basement would be entrained, which the licensee contends should not cause anincrease in the content of radioactive materials in the pool water above the levels experienced while the reactor was operating. The licensee concludes that the resulting levels of radioactive material from the evaporation of the water spread over the basement and first floors would be limited to the tritium contained in the water and would be less than the evaporation ratesexperienced from the surface or the reactor pool while the reactor was operating. The otherradionuclides would remain in the facility. The licensee stated that during normal operation ofthe reactor pool, prior to reactor shutdown, the 240 square feet of the pool's surface was maintained between 90 F and 116 F with an estimated evaporative loss rate of 4 gallons perhour.3.2.5.3 Tritium-Loaded Heavy-Water Spill The consequences of a spill from a 55-gallon drum of tritium-loaded heavy water would be theemission of tritium via the FNR exhaust stack. Taking credit for the FNR exhaust stack dilution factor of 400, and assuming the emission of 8000 cubic feet per minute up the FNR exhaust stack, the licensee calculated the emission of tritium from the facility to be 9.1 mCi per hourbased upon the AEC limit for a full year as specified in Table 2 of Appendix B to 10 CFR Part 20. The heavy-water reflector contains the most concentrated tritium-loaded heavy water. At 217 Ci of tritium (as of April 2004) in an estimated 50 gallons, the licensee calculated the highest estimated concentration of tritium to be 1.1 mCi/milliliter (mL). Given thisconcentration, a spill from this tank would require the evaporation rate to be limited toapproximately 9 mL per hour if the emission were averaged over an entire year. Any spill oftritium-loaded heavy water could be easily flushed to the floor drains for collection in the hot and cold sumps and eventual collection in the retention tanks. Conservatively, 1 week or less would be needed to clean up the spill or to stop the tritium evaporation. The licensee calculated thatthe emission rate could increase to 473 mCi/hour over 1 week. This equates to an evaporation rate of 0.473 liters (L) of the tritium-loaded heavy water per hour for the entire week of cleanup activities. The licensee concludes that the emission of tritium at a rate of 473 mCi per hour for the 1 week of cleanup would result in a maximum exposure to an individual of 50 mrem.In the event of a spill of the heavy-water reflector while still in the reactor pool, the licenseestated that the dilution by the water in the pool would decrease the concentration of the tritiumin the water source and result in a lower emission rate of tritium from the facility (see LicenseAmendment Nos. 36 and 46).3.2.5.4 Conclusions The licensee analyzed bounding accidents that may occur during the decommissioning project. Based on the NRC staff's review of the information provided by the licensee, the radiologicalconsequences for the types of accidents that may potentially occur during decommissioning ofthe FNR are bounding and within the limits specified in 10 CFR Part 20. 3.3 Decommissioning Activities 3.3.1 Radiological Status of the Facility 3.3.1.1 General The licensee listed potential causes of radioactive contamination in the reactor building fromnormal operations and routine activities as well as nonroutine occurrences, operations, accidents, and spills. Based on these historical reviews, in addition to characterization surveysof the facility during nearly 50 years of operation of the FNR, the licensee determined thatevents occurred that led to the radiological contamination of the facility. However, the prompt response and cleanup activities initiated by the facility staff limited contamination in areas notexpected to be contaminated by routine operations. Additionally, the licensee's practice ofperiodic monitoring and maintaining contamination action levels between 3 and 10 times background resulted in a limited number of areas where contamination levels are reported to exist above the anticipated release criteria.3.3.1.2 Principal Radioactive Components The information obtained by the licensee indicates that the radioactive portions of the facility areprimarily confined to the reactor internals and reactor pool. The licensee estimated the radioactivity inventory by considering the constituent elements of the material in question and calculating the duration of exposure to the neutron flux and the energies of the incidentneutrons. The licensee is responsible for performing direct measurements during actual removal and/or dismantlement of components. Those data will be used as the basis forspecifying the necessary safety measures and procedures to maintain exposures at ALARA levels during the various dismantlement, removal, decontamination, and waste packaging and storage operations.3.3.1.2.1 Pool Water The reactor pool and the primary cooling system contain approximately 50,000 gallons of waterrequiring removal. The water was supplied from potable water through filters and demineralizers. The cleanliness of the water was maintained by a system of filters, and H-OHdemineralizers were used as necessary to maintain the conductivity to less than 5 micro-ohm per centimeter. Chemical additions to the water were not required to maintain the pH between 4.5 and 7.5. Table 3-2 lists the levels of radioactivity in the pool water measured in March

2004.Table 3-2 Radioactivity of the Reactor Pool Water (March 17, 2004)

Gross Alpha<7.18 pCi/lGross Beta699 pCi/l Tritium1,110,000 pCi/l Silver-108m66.8 pCi/l Silver-110m1,150 pCi/l Zinc-65645 pCi/l3.3.1.2.2 Bridge Suspension Frame and Grid PlateThe reactor grid plate has overall dimensions of approximately 25 inches by 33 inches by 6inches thick. Figure 3-2 illustrates the results of a survey of the reactor grid conducted by thelicensee, using an Eberline RO-7 with a high-range probe in an underwater housing. The licensee believes a large contribution to the dose rates for the interior positions of the reactor grid, two rows in from each edge, originates from the 0.25 inch by 1 inch stainless steel alignment pins at these locations. Additionally, 18 stainless steel bolts (0.25 inch by 5 inches) are present around the outer edge of the reactor grid and were used to suspend the hopper from the bottom of the reactor grid.

Figure 3-2 Radiation Levels (R/h) on the Reactor Grid Plate (April 2004) 3.3.1.2.3 Heavy-Water ReflectorAccording to the licensee, the last transfer of heavy water to the heavy-water reflector occurredin February 1992. Subsequently, heavy-water transfers to maintain the tritium content of the heavy-water reflector below 50 Ci were no longer required because of the removal of the 50-Ci limit (see Amendment 36 to the TSs). The licensee assumed that the reflector contained the maximum allowed activity of 50 Ci of tritium prior to the transfer and contained 44.6 Ci of tritium after completion of the 5-gallon transfer. The licensee calculated the annual tritium buildup in the heavy-water reflector using an average tritium production rate and the power history for each year and accounting for radioactive decay. The licensee evaluated the tritium inventory in the heavy-water reflector at the end of March 2004 to be 217 Ci. The heavy water is on loan from DOE and will be returned to the Savannah River Site.3.3.1.2.4 Beam Ports The reactor staff have or will remove the collimators or experiment plugs that were installed inmost of the 6-inch and 8-inch beam ports, with the following exceptions:*The upper 6-inch through port, running east-west is believed to contain shielding materialsand a 0.75-inch aluminum tube that runs through this shielding material and is open on the east and west ends of the through port. The contents of this through port will requiredetailed characterization before removal, processing, and disposal*The beam port closest to the thermal column is believed to contain a collimator that ext endsthe full length of the beam port (i.e., from the opening to the reactor core). This collimator isbelieved to contain a 0.5-inch stainless steel box in a taper, small at the reactor end and wide at the opening, around which lead and polyethylene shielding was attached. The contents of this beam port will require detailed characterization before removal, processing,and disposal. Dose rates of several R/h or higher are expected from the reactor end of this collimator. A dose rate of approximately 35 mrem per hour is present at the open end of thecollimator, which extends from the beam port opening.A 500-pound lead door or shutter shields each 6-inch beam port opening, and a 630-poundlead door shields each 8-inch beam port opening. These doors can be raised or lowered across the beam port opening (for the lower 6-inch through port, the door moves side to side rather than up and down).The licensee stated that the contents of all remaining beam ports have been or will be removedand concrete port plugs reinstalled. Based upon its past experience, the licensee expects minimal contamination and little or no activation from these items.3.3.1.2.5 Thermal Column The licensee recently opened the thermal column and compared it with the facility drawings. This investigation by the licensee indicated that half of the graphite was removed in the late1960s to early 1970s. Surveys of the graphite bl ocks showed elevated levels of contaminationin areas where water had calcified, but radiation exposure levels were indistinguishable from background. Small samples of some of the graphite material were taken from the graphite blocks in the approximate center of the column, and results of the analysis of these samples are pending. From facility drawings, the thermal column contains the volumes of materialslisted in Table 3-3 below.The licensee stated that the extent the materials in the thermal column are surfacecontaminated or activated is not known. The licensee will perform detailed characterization before and after removal to determine the options for disposal.Table 3-3 Estimated Material Volumes for the Thermal Column MaterialVolume (ft

3) or Mass (lb)Cadmium sheet0.8 ft 3Boral (boron carbide between aluminumcladding)1.0 ft 3Graphite block96 ft 3Lead shot3,752 lbLead block16,747 lb Other lead (caulk and thin strip)50 lb3.3.1.2.6 Pneumatic Tube SystemSix of the eight tubes to the irradiation stations on the west side of the reactor grid are currentlyplugged at the point where the tubes penetrate the floor of the reactor pool. The licensee willperform detailed characterization before removal, processing, and disposal.3.3.1.2.7 Other Items in the Reactor Pool The following are some of the other, higher level, radioactive components to be handled andprocessed during FNR decommissioning based on process knowledge and direct measurements performed by the licensee:*reactor irradiation facility for large samples (RIFLS) reading as high as 50 R/hr*heavy section steel irradiation (HSSI) experiment reading approximately 11,200 R/hr
  • reactor control/shim rods reading about 2500 R/hr3.3.1.3 Sanitary Sewer Lines According to the licensee, from the opening of the PML in 1954 until the summer of 1991,liquids containing low levels of radioactivity were discharged from the retention tanks in the PML to the sanitary sewer line following sampling to verify that applicable regulatory limits and license conditions were satisfied. The sanitary sewer line runs south along the western side of the PML, turns west and follows Bonisteel Boulevard towards the UM hospital, at which point it turns and runs along the river to the Ann Arbor sewage treatment plant. The licensee collected a sample of the internal pipe surface of the sewer line at the point it exits the PML. Attempts bythe licensee to obtain sludge (solids) from several locations along this pathway were unsuccessful, given the small volume of sludge that was present at most locations. Thesamples, therefore, mainly consisted of liquid. There was no detectable radioactivity from the PML. 3.3.1.4 Soil beneath the Reactor Building Because of the possibility of pool water leakage and the 1993 loss of approximately 7500gallons of low-level radioactive water, the licensee conducted an investigation to assess whether soil underlying the reactor pool around the reactor building was contaminated. Duringthe investigation, the licensee performed soil borings (1) immediately north of the reactor poolthrough the first floor into the unexcavated area, (2) through the basement floor near the pointwhere the foundation drain tile connects to the cold sump (the source of the 7500-gallon leak in 1993), and (3) immediately east of the drain-tile line just outside the reactor building. This investigation detected only ambient levels of radionuclides normally present in soil, andtritium at a concentration of 14.5 pCi per gram in the upper foot of soil located immediatelynorth of the reactor pool. 3.3.1.5 Ground Water Because of the 7500 gallons of low-level radioactive water released in 1993, and the possibilityof leaks from the reactor pool itself, the licensee investigated the potential for radionuclides inthe ground water near the reactor building. Since a previous monitoring well in this location was decommissioned because it dried up, the licensee established a new ground water monitoring well in April 2003 immediately south of the PML. Sampling of this well found, with the exceptionof 333 pCi/L of tritium (well below the U.S. Environmental Protection Agency maximum contaminant level of 20,000 pCi/L), no detectable radionuclides present in ground water otherthan those present in background water samples.3.3.1.6 Radionuclides According to the licensee, radionuclides expected to be encountered during decommissioningof the FNR originated from reactor operations as well as experiments performed over the years.

Several of these radionuclides have short half-lives. The licensee determined the potential radionuclides present, shown in Table 3-4, through research of FNR historical documents and interviews with knowledgeable personnel.During characterization of accessible areas, the licensee identified cobalt-60 and cesium-137as the dominant contaminants with smaller amounts of numerous other activation and fission products. As such, the licensee concluded that the radionuclide mix does not appear to be uniform. Table 3-4 List of Potential Radionuclides NuclideHalf-Life(yr)DecayModeNotesAntimony-1252.8-, AP; from n-activation of materials containingtinBismuth-210m3.0x10 6, AP; from n-activation of SS hardwareCadmium-1091.26 , AP; from n-activation of cadmium metal ormaterials containing cadmiumCarbon-145.73x10 3-AP; from n-activation of graphite or materialscontaining carbonCesium-1342.1-, AP; from n-activation of cesium, FP; minor FPinventory constituentCesium-13730.2-, FP; expected to be predominant FP speciespresentCobalt-605.3 , -, +, AP; from n-activation of SS hardware;expected to be predominant AP species presentEuropium-15213.5-, AP/FPEuropium-1548.5-, FPIron-552.7AP; from n-activation of SS hardware ormaterials containing ironManganese-540.86 , AP; from n-activation of SS hardwareNickel-597.5x10 4 , AP; from n-activation of SS hardwareNickel-63100-AP; from n-activation of SS hardwareScandium-460.23-, AP; from n-activation of materials used intesting/experimentsSilver-108m127 , AP; from n-activation of materials containingsilverSilver-110m0.68-, AP; from n-activation of materials containingsilver (Ag-110m)Tritium12.3-AP; from n-activation of water and from shield tankZinc-650.67 , +, AP; from n-activation of SS hardware - = beta, + = positron, = electron capture, = gamma ray Note: The list of potential radionuclides provided above is based on the assumption that operations ofthe FNR have resulted in the neutron activation of reactor core components and other integral hardwareor structural members that were situated adjacent to, or in close proximity to, the reactor core duringoperations. Specific items that are considered to have been exposed to neutron activation includematerials composed of aluminum, steel, stainless steel, graphite, cadmium, lead, concrete, and possiblyothers. Neutron activation of materials beyond the concrete liner/biological shield structure (i.e., intosurrounding soil volumes) is not expected for the FNR based on earlier studies, experience from similarresearch reactor decommissioning projects, reactor-specific calculations that considered measuredvalues for neutron leakage fluence, integrated operating power histories, reactor core/pool structuralconfigurations, and material composition of pool structures.

3.3.1.7 Conclusions The staff has reviewed the dose rates and contamination levels identified by the licensee andthe licensee's plans for followup surveys. Based on experience and professional judgment, the staff concludes that the licensee's estimates of the radiological conditions and radiation measurements are acceptable. The staff finds that a followup characterization survey will be necessary following the removal of the material from the pool and pool draining. This survey will include the pool and any leakage pathways. Based on review of the information providedby the licensee, the staff concludes that the radiological status of the FNR has been adequatelycharacterized and that this facility can be safely decommissi oned.3.3.2 Radiological Release Criteria The licensee proposed the DECON decommissioning alternative for the reactor. The licenseestated that the results of the site and facility radiological characterization survey indicate thatthe building structures may not need extensive decontamination to meet the release criteria.The licensee proposed that the FSS will use derived concentration guideline levels (DCGLs)developed from the characterization survey data and the current NRC guidance for licensetermination in 10 CFR Part 20. The regulations in 10 CFR 20.1402 allow termination of a license and release of a site for unrestricted use if the residual radioactivity that is distinguishable from background radiation results in a TEDE to an average member of a critical group of less than 25 mrem (0.25 millisievert) per year, and the residual radioactivity has beenreduced to ALARA levels.3.3.2.1 Structure Surfaces The licensee proposed that for remediation activities, it will select the DCGLs for resi dualradioactive material contamination on FNR structural surfaces from the tables of NRC default screening values (refer to NUREG-1757, "Consolidated NMSS Decommissioning Guidance"). Table 3-5 lists the screening values for total structure surface contamination; guideline levels for removable activity are 10 percent of the values in the table. The NRC has conservatively evaluated these default screening levels as satisfying the goal that doses to facility occ upantsand the public during future facility use do not exceed 25 mrem annually. Default screeningcriteria are based on conservative exposure scenario and pathway parameters and are generally regarded as providing a high level of confidence that the annual dose limits will not be exceeded. These screening values are applicable where it can be demonstrated that theresidual radioactivity is present on the surface only and volumetric contamination (less than 10 millimeters deep) is not present.Table 3-5 Acceptable License Termination Screening Values of Common Radionuclidesfor Structure Surfaces RadionuclideSymbolAcceptable ScreeningLevels1,2 for UnrestrictedRelease (dpm/100 cm 2)3Tritium 3H1.2E+08Carbon-14 14C3.7E+06Sodium-22 22Na905E+03Sulfur-35 35S1.3E+07Chlorine-36 36Cl5.0E+05Manganese-54 54Mn3.2E+04Iron-55 55Fe4.5E+06Cobalt-60 60Co7.1E+03Nickel-63 63Ni1.8E+06Strontium-90 90Sr8.7E+03Technetium-99 90Tc1.3E+06Iodine-129 129I3.5E+04Cesium-137 137Cs2.8E+04Iridium-192 192Ir7.4E+04 Notes: 1Screening levels presented here are taken from the NRC's "Supplemental Information on theImplementation of the Final Rule on Radiological Criteria for License Termination," issued 1998. The DP states that the licensee will develop site-specific screening levels for the project in themanner described in that reference. 2Screening levels are based on the assumption that the fraction of removable surface contaminationis equal to 0.1. For cases in which the fraction of removable contamination is undetermined orhigher than 0.1, users may assume for screening purposes that 100 percent of the surfacecontamination is removable, and therefore the screening levels should be decreased by a factor of10. Users may calculate site-specific levels based on available data on the fraction of removablecontamination and DandD version 2. 3Units are dpm/100 cm 2; 1 dpm is equivalent to 0.0167 becquerel (Bq). Therefore, to convert tounits of Bq/square meter (m 2), multiply each value by 1.67. The screening values represent surfaceconcentrations of individual radionuclides that would be deemed in compliance with the 0.25millisievert per year (mSv/yr) (25 mrem/yr) unrestricted release dose limit in 10 CFR 20.1402. Forradionuclides in a mixture, the "sum of fractions" rule applies (see Note 4 in Appendix B to10 CFR Part 20).Characterization surveys performed by the licensee have identified multiple radionuclidecontaminants on surfaces and in various media at the FNR. Predominant contaminantsanticipated by the licensee at the time of license termination are cobalt-60 and cesium-137. However, additional fission and activation products are present on some surfaces, generally at lower concentrations and at spotty distributions. The licensee described that, for surfaces, it willdetermine concentrations of specific contaminants and ratios to their respective DCGLs to demonstrate satisfaction of the Unity Rule as described in Section 4.3.3 of the MARSSIM (Ref. 9). The licensee will use gross beta measurements to demonstrate compliance with surface activity guidelines, and it will base the gross beta DCGL on measurements of surr ogatecontaminants with known relationships to the total contamination mix.The DCGLs described are net (above background) concentrations and activity levels ofradionuclides; the licensee will make appropriate adjustments for instrument background levelsand naturally occurring radionuclide concentrations in various media before comparing data to the respective DCGLs.Because of the conservatism used in the development of the default screening values, furtherevaluations and actions are not required to reduce residual radioactivity to ALARA levels.3.3.2.2 Surface Soil and Sediment The licensee proposed that for remediation activities, it will select the DCGLs for resi dualradioactive material contamination in sediments or surface soil (top 15 cm of soil) under or near the FNR from the tables of NRC default screening values (refer to NUREG-1757). Table 3-6lists the screening values for contaminants in soil. These default screening levels provide assurance that doses to facility occupants and the public during future facility use do notexceed 25 mrem annually. These default screening criteria are based on conservative exposure scenario and pathway parameters and are generally regarded as providing a high level of confidence that the annual dose limits will not be exc eeded.Table 3-6 Acceptable License Termination Screening Values of Common Radionuclidesfor Surface Soil RadionuclideSymbolSurface ScreeningValues1,2Tritium 3H1.1E+02Carbon-14 14C1.2E+01Sodium-22 22Na4.3E+00Sulfur-35 35S2.7E+02Chlorine-36 36Cl3.6E-01Calcium-45 45Ca5.7E+01Scandium-46 46Sc1.5E+01Manganese-54 54Mn1.5E+01Iron-55 55Fe1.0E+04Cobalt-57 57Co1.5E+02Cobalt-60 60Co3.8E+00Nickel-59 59Ni5.5E+03Nickel-63 63Ni5.5E+03Strontium-90 90Sr1.7E+00Niobium-94 94Nb5.8E+00Technetium-99 99Tc1.9E+01Iodine-129 129I5.0E-01Cesium-134 134Cs5.7E+00Cesium-137 137Cs1.1E +/- 01Europium-152 152Eu8.7E +/- 00Europium-154 154Eu8.0E +/- 00Iridium-192 192Ir4.1E +/- 01Lead-210 210Pb9.0E-01Radium-226 226Ra7.0E-01 RadionuclideSymbolSurface ScreeningValues1,2-51-Radium-226+C3226Ra+C6.0E-01Actinium-227 227Ac5.0E-01Actinium-227+C 227Ac+C5.0E-01Thorium-228 228Th4.7E+00Thorium-228+C 228Th+C4.7E+00Thorium-230 230Th1.8E+00Thorium-230+C 230Th+C6.0E-01Thorium-232 232Th1.1E+00Thorium-232+C 232Th+C1.1E+00Protactinium-231 231Pa3.0E-01Protactinium-231+C 231Pa+C3.0E-01Uranium-234 234U1.3E+01Uranium-235 235U8.0E+00Uranium-235+C 235U+C2.9E-01Uranium-238 238U1.4E+01Uranium-238+C 238U+C5.0E-01Plutonium-238 238Pu2.5E+00Plutonium-239 239Pu2.3E+00Plutonium-241 241Pu7.2E+01Americium-241 241Am2.1E+00Curium-242 242Cm1.6E+02Curium-243 243Cm3.2E+00Notes: 1These values represent surface soil concentrations of individual radionuclides that would bedeemed in compliance with the 0.25 mSv/yr (25 mrem/yr) unrestricted release dose limit in 10 CFR20.1402. For radionuclides in a mixture, the "sum of fractions" rule applies (see Note 4 inAppendix B to 10 CFR Part 20).2Screening values are in units of pCi/g equivalent to 0.25 mSv/yr (25 mrem/yr). To convert frompCi/g to units of Bq per kilogram (Bq/kg), divide each value by 0.027. These values were derivedusing DandD screening methodology (NUREG/CR-5512, Volume 3, "Residual RadioactiveContamination for Decommissioning"). They were derived based on selection of the 90thpercentile of the output dose distribution for each specific radionuclide (or radionuclide with the specific decay chain). Behavioral parameters were set at "Standard Man" or at the mean of thedistribution for an average human. 3"Plus Chain (+C)" indicates a value for a radionuclide with its decay progeny present in equilibrium. The values are concentrations of the parent radionuclide but account for contributions from thecomplete chain of progeny in equilibrium with the parent radionuclide (NUREG/CR-5512, Volumes1, 2, and 3).

Characterization surveys performed by the licensee have identified multiple radionuclidecontaminants at the FNR that could also be present in soil and sediment. Predominantcontaminants anticipated by the licensee are cobalt-60 and cesium-137. However, additionalfission and activation products could also be present in soil and sediment. In addition, the licensee described that variable radionuclide mixtures may be present in soil and sediment.

Therefore, the licensee will determine concentrations of specific significant contaminants andratios to their respective DCGLs in a manner satisfying the Unity Rule, as described in Section 4.3.3 of the MARSSIM (Ref. 9).The criteria described are net (above background) concentrations and activity levels ofradionuclides; the licensee will make appropriate adjustments for instrument background levelsand naturally occurring radionuclide concentrations in various media before comparing data to the respective criteria. Because of the conservatism used in establishing the default screening values, furtherevaluations and actions to demonstrate that the final conditions satisfy ALARA provisions are not required.3.3.2.3 Subsurface and Inaccessible Structures The criteria for residual radioactive contamination on FNR facility surfaces discussed in Section3.3.2.1 are not applicable for surfaces where the contaminant is not at the surface (greater than10 millimeters deep), activated surfaces, and inaccessible areas excluding buried pipes, etc. The licensee must still develop the criteria for radioactive contamination of these types of surfaces because it has not yet obtained characterization results. However, it will develop thespecific release criteria that will be applied in these instances at a later date using RESRAD-BUILD or equivalent methodology. The licensee will develop the criteria to ensure thatestimated doses to facility occupants and the public during future facility use is less t han 25mrem annually. Characterization surveys performed by the licensee have identified multiple radionuclidecontaminants on surfaces and in various media at the FNR. Predominant contaminantsanticipated by the licensee at the time of proposed license termination are cobalt-60 andcesium-137. However, additional fission and activation products are present in some media, generally at lower concentrations and at spotty distributions. The licensee described that variable radionuclide mixtures are also present for different media. The licensee will determineconcentrations of specific significant contaminants and ratios to their respective DCGLs in a manner satisfying the Unity Rule, as described in Section 4.3.3 of the MARSSIM (Ref. 9).The License Termination Rule (10 CFR Part 20, Subpart E, "Radiological Criteria for LicenseTermination") also requires that residual radioactivity resulting from licensed material for release to unrestricted use must be at ALARA levels. The licensee may need to further reduce thecriteria for residual radioactive material contamination of subsurface structures or components within the physical structure of the FNR facility (left after remediation or decontamination) to satisfy the ALARA requirement. Reduction of the cleanup criteria for subsurface and inaccessible structures may follow an examination by the licensee of the reduction in the estimated dose to the facility occupants and the public using the RESRAD-BUILD softwarecombined with an examination of the costs associated with achieving these reduced levels of residual radioactivity. The licensee will document this evaluation in its final report to the NRC.The criteria described in this section should be net (above background) concentrations andactivity levels of radionuclides; the licensee will make appropriate adjustments for instrumentbackground levels and naturally occurring radionuclide concentrations in various media before comparing data to the respective criteria.3.3.2.4 Conclusions The licensee has adequately specified the radiological release criteria need for licensetermination that will be used for accessible building surfaces and soil. The staff concludes thatthe licensee understands the release criteria for license termination for the FNR and has proposed acceptable DCGLs in accordance with applicable guidance.3.3.3 Decommissioning Tasks 3.3.3.1 Characterization Surveys The licensee has conducted characterization studies as part of the planning activities for theDP. The licensee has identified the type, quantity, condition, and location of radioactive and/or hazardous materials that are or may be present in the FNR. It conducted extensive surveys of accessible areas of the FNR in September 2002 and April 2003. The characterization report provided in Appendix A to the DP summarizes the results of these surveys. The licensee willperform additional surveys in conjunction with the dismantlement and decontamination activities discussed below, as previously inaccessible areas are made accessible.3.3.3.2 Dismantlement and Decontamination of the FacilityDismantling and decontamination will be required to remove materials that were activated orradiologically contaminated during operation of the FNR in order to meet the unconditionalrelease criteria for license termination. The licensee will employ standard industry dismantlingand decontamination techniques using tools such as wire saws, high-pressure/ultra high-pressure water, needle guns, jack hammers, torches/plasma arc torches, hydraulic cutters, and hand tools, following approved procedures or work packages. The following sections discuss typical dismantling and decontamination activities. The licensee may opt not to follow the sequence for ALARA, safety, accessibility, or scheduling reasons.3.3.3.2.1 Systems Formerly Important to Safety As decommissioning progresses, the licensee may inactivate (deenergize and isolate) orremove all inactive systems or systems not currently required by the TSs or laterdecommissioning activities but formerly identified in the SAR. The licensee has identified several systems, structures, or components that will be removed from the facility in accordancewith the change control process defined in 10 CFR 50.59 and Section 9.0 of the DP, including the following:*standby generator*heavy-water reflector

  • spent fuel storage racks
  • pneumatic t ube system external to the reactor pool*secondary cooling system*emergency cooling system
  • control console
  • exhaust for hood in Room 3103
  • exhaust for pneumatic blowers, first-floor trunks around pool, and storage ports
  • beam port extensions3.3.3.2.2 Other Systems Systems identified by the licensee that may be deenergized and/or isolated include the potablewater line, drain lines to the hot or cold sump, reactor air to miscellaneous supplies, gaseous nitrogen supply lines, and the demineralized water supply to the PML. Systems interfacing with the contiguous wall of the PML will be isolated on the PML side of the interface, when practical. The licensee will apply the quality assurance requirements identified in Section 1.3.4.2 of theDP when required.3.3.3.2.3 Asbestos The licensee will remove, package, and dispose of radioactively contaminated asbestos-containing materials in accordance with applicable regulations. It may also remove, survey, and dispose of uncontaminated asbestos-containing materials in accordance with applicable regulations.3.3.3.2.4 Temporary Systems The licensee may need to install temporary systems to support decommissioning activities. These may include additional electrical outlets for temporary ventilation or decontamination equipment, a water purification system to purify or decontaminate liquids, openings in thereactor building for equipment access or waste removal, waste storage and handling systemsor equipment, service air, potable waste, fire detection, and fire hose stations.3.3.3.2.5 Reactor Pool The licensee will estimate the radioactivity associated with the high-dose items (the reactor grid,shim and control rods, beam port extensions, etc.) when these items become accessible. The licensee will reduce the size of the reactor grid plate, shim and control rods, heavy-waterreflector, pneumatic tubes, RIFLS, HSSI experiment, and remaining miscellaneous high-dose items to facilitate loading into HICs or, for inherently stable items, a liner. To do this, thelicensee may use long-reach tools, remotely operated equipment, human divers, or a combination of these techniques. The licensee plans to use the water in the reactor pool as shielding and for contaminationcontrol during high-dose item size reduction and removal activities. However, it may be necessary for the licensee to lower the water level or drain the pool to remove items such as the pneumatic tube bundle penetration that could, upon removal, introduce a potential pooldrainage pathway. If the pool water levels are lowered, the licensee may use shielding or remote size reduction techniques to maintain personnel exposure at ALARA levels.The licensee may transfer high dose-rate items, such as the shim and control rods, RIFLS, andHSSI experiment, to the hot cells in the PML for size reduction. High dose-rate items may also be transferred and loaded dry into the HIC or liner using shields.Once the high dose-rate items are loaded into the HIC or liner, the licensee will place the HIC orliner into an approved, shielded shipping cask for transport to an approved disposal site. The HIC or liner should be directly loaded into a shipping cask submerged in the reactor pool (similar to the methods when loading and shipping irradiated fuel elements), whenever the size of the cask permits. The licensee recognizes that the HIC or liner may require indirect loading using a shielded transfer cask if the size of the cask or other factors prohibits loading in the reactor pool.The licensee will dispose of the water in the reactor pool when the water is no longer useful asa radiological shield or for contamination control. The licensee will filter and treat the liquid fromthe pool and piping as necessary to meet discharge requirements of the license as well as Federal, State, and local laws. Liquid effluents will subsequently be discharged to the City ofAnn Arbor sanitary sewer using approved procedures. The licensee will treat, stabilize, andpackage liquids not meeting release criteria to meet shipping requirements and waste acceptance criteria at an approved disposal site.Following draining of the pool, the licensee will characterize the structure to determine theextent and depth of activation and contamination in the reactor pool floor, walls, and embedded beam port tubes. UM may use the characterization results to select either the pool removal or pool decontamination option for decommissioning based on ALARA, safety, structural, cost, schedule, and future use considerations.Future licensee plans for the reactor building require the decontamination and removal of thereactor pool from the building. UM has elected to remove those portions of the reactor pool that may not be readily remediated. Contingent upon the results of the reactor poolcharacterization, the reactor pool walls and possibly portions of the reactor pool floor will be cutinto large bl ocks and packaged and shipped as radioactive waste by the licensee to a licenseddisposal facility. The licensee may not remove materials embedded in the concrete (beam porttubes, drain pipes, conduit, tile, etc.) unless it is necessary to meet transportation requirements and the disposal site waste acceptance criteria.If decontamination of the reactor pool or a portion of the reactor pool is elected for thedecommissioning option, then the licensee will decontaminate pool surfaces and the activatedconcrete will be removed to levels that will facilitate termination of the license. The licensee willcollect core bore samples to evaluate subsurface contamination. The licensee stated that contamination present below surfaces (e.g., surface cracks or voids) will be decontaminated or removed, and the waste generated will be packaged and shipped to a licensed disposal site. 3.3.3.2.6 Embedded PipesThe licensee will decontaminate or remove contaminated pipes, drains, and conduit embeddedin concrete. Sludge, scale, and other waste generated will be treated or stabilized andpackaged to meet the disposal site waste acceptance criteria. The licensee will dischargedecontamination liquid to the sanitary sewer if it meets the license requirements as well as Federal, State, and local requirements for discharge to the sewer. 3.3.3.2.7 Surface and Subsurface Sampling The licensee proposed to collect sufficient soil samples from unexcavated areas beneath andwest of the pool to determine if an unknown leak in the pool contaminated the soil surroundingthe pool. The licensee will seal or plug any holes drilled through the concrete to prevent thehole from becoming a potential pathway to the environment.3.3.3.2.8 Contaminated EquipmentThe licensee will remove or decontaminate contaminated equipment from each floor of theFNR. Essential equipment, such as heating, ventilation, and air conditioning (HVAC) and electrical and instruments interfacing with PML or FNR systems, may be isolated to reduce thepotential for accidental releases of water or energy. Examples of equipment that may need to be decontaminated or removed include the following:*basement-primary coolant piping and instrumentation, holdup tank, primary pump andmotor, ion exchange pipi ng and system, and hot and cold sump pumps and motors*first floor-HVAC ducts, source storage ports, transfer chute, thermal column and thermalcolumn door, and drain lines and piping not embedded in concrete*second floor-HVAC equipment, ducts and butterfly valves
  • third floor-reactor bridge, remaining reactor suspension frame, pool and reactorinstrumentation, heavy-water reflector support equipment, HVAC, drain lines and piping, pool filter/vacuum system, and any miscellaneous low-dose items in or attached to the pool*fourth floor-crane over the pool, HVAC (contamination not expected in all components) 3.3.3.2.9 Remaining Areas The licensee will decontaminate or remove any remaining contaminated areas within the FNR,then survey to confirm the area has been decontaminated to levels that will meet unconditionalrelease criteria. Examples of areas that may require decontamination include the following:*basement-concrete floor, hot and cold sumps, holdup tank pit, ion exchange pit, and walls
  • first floor-floor, wall by the source storage ports, and thermal column door trench
  • third floor-laboratories, floor around the pool, and the south wall The licensee does not expect decontamination to be required on the second and fourth floors. The licensee will package and dispose of waste generated during this activity at a licenseddisposal site.3.3.3.2.10 Soil and Buried Pipe Remediation If contaminated soil is identified and the source of the contamination is the FNR, the licenseewill evaluate the results against the release criteria. If contamination levels require soil removal,the licensee will remove, package, and dispose of the soil at an approved disposal site.The licensee will package and dispose of any buried pipes (e.g., drain tiles) found to beradiologically contaminated that cannot be decontaminated on site to meet final release criteria. The licensee will collect final release samples after remediation of the soil or buried pipes. However, the excavations will remain open to permit the NRC to perform confirmatory surveysor sampling. The licensee will collect split samples before backfilling if backfilling is necessary for safetyreasons before confirmatory surveys are performed. The licensee will notify the NRC of the expected completion date of the remediation so that the NRC has the opportunity to be presentto verify collection of soil samples. Once NRC concurrence is received, the licensee will backfill the excavation to reduce any potential safety hazard.The assumption that neutron activation of the soil beneath the reactor pool did not occur will beconfirmed by evaluating the activation of the concrete floor in the void directly beneath the reactor core, which is accessible from the reactor basement (refer to Figures 2-2 and 2-7).3.3.3.3 Final Survey and Report Following decontamination and remediation activities of the FNR, the licensee will perform afinal radiological survey covering the entire FNR. A final radiological survey, executed according to the approved FSS plan, will document that the licensee's decommissioning effortsachieve the release criteria.Once all decontamination has been performed and verified through final radiological surveys,the licensee will develop a final release report. The licensee will record in this report thedecontamination and remediation activities performed and document the final radiological status of the FNR facility and associated grounds. The licensee will use this final report in partas the basis of the application for license termination.3.3.3.4 Conclusions Based on review of the information provided by the licensee, the plans for decommissioning theFNR facility follow an acceptable sequence and are acceptable to the NRC staff.3.3.4 Schedule The scheduled time from regulatory approval of the DP to the request for release of the site forunrestricted use is estimated to be 15 months. The licensee proposed that changes to the schedule may be made at the discretion of UM, including changes due to resource allocation, availability of a radioactive waste burial site, interference with ongoing UM activities, ALARAconsiderations, further characterization measurements, and/or temporary onsite radioactive waste storage operations.3.3.4.1 Conclusions Based on a review of the licensee's proposed decommissioning schedule, the staff concludesthat the licensee's proposed schedule is acceptable.3.3.5 Proposed Final Status Survey Plan The licensee provided a plan for the development, review, and approval of the FSS plan oncethe site is fully characterized. The licensee's stated objective of the FSS is to ensure that the facility meets the unrestricted release criteria.3.3.5.1 General Survey Approach The licensee noted that all factors influencing the FSS for the FNR are not available and will notbe available until it evaluates more facility details following additional characterization activitiesto be conducted upon approval of the DP. The outline for the proposed FSS plan prepared by the licensee is intended to provide information to the NRC for determining the adequacy of thelicensee's understanding of the proposed FSS plan as it pertains to the goal of remediation in a manner satisfying the radiological criteria for license termination. The final FSS plan, which the licensee will formally submit for approval at a later date (included as a license condition; refer toSection 4.0), will adequately demonstrate compliance with the radiological criteria for licensetermination. The licensee prepared its proposed FSS plan in accordance with the guidelines andrecommendations presented in the MARSSIM (Ref. 9). The licensee committed to implement the MARSSIM process that emphasizes the use of data quality objectives (DQOs) and data quality assessment, along with a quality assurance and quality control program. As such, the licensee will follow the graded approach concept of the MARSSIM to assure that survey effortsare maximized in those areas having the greatest potential for residual contamination.The licensee committed to conducting the FSS with trained radiological control technicians, whofollow standard, written procedures and use properly calibrated instruments, sensitive to the potential contaminants. The licensee may develop designs for specific surveys for some areas, including determinationof specific nuclide mixture guidelines, sampling or measurement methods, survey unit identification and classification, and data evaluation techniques, at the time of the survey inaccordance with the guidance presented in the proposed FSS plan.3.3.5.2 Instrumentation The licensee will base the selection of instruments on the type of radiation emitted for the radionuclides of interest, as well as the required range, accuracy, and tolerance needed todemonstrate conformance to specified requirements. Selection and use of instrumentation for the FSS will also be based upon the need to ensure that the residual radioactivity remaining onsite meets the release criteria. Table 3-7 lists the instrumentation the licensee intends to usefor the FSS and associated documentation (e.g., characterization information used in the design of the final survey), along with estimated detection sensitivities. The licensee will alsoaccept other instruments that are the functional equivalent of those listed.Table 3-7 Instrumentation for FNR Radiological Surveys DetectorTypeMakeMeterApplicationSensitivity (dpm/100 cm 2, except as noted)ScanningStatic Count(1 minute)43-68GasProportionalLudlum2221Gross beta scan andmeasurement120050043-68GasProportionalLudlum2221Nickel-53 Gross betascan andmeasurement5000200043-67FloorMonitorLudlum2221Gross beta scan800N/A43-68GasProportionalLudlum2221Gross alphameasurement20070TennelecLB5100GasProportionalTennelecN/AGross alpha smearmeasurementN/A5TennelecLB5100GasProportionalTennelecN/AGross beta smearmeasurementN/A1044-10NaILudlum2221Gamma scan10 pCi/gN/ABecause the radionuclides expected by the licensee to be present as contaminants emit (withfew exceptions) beta particles with maximum energies greater than 0.300 megaelectron volts (MeV), detector efficiencies for measuring surface activity are generally determined using technetium-99 (maximum beta energy of approximately 0.292 MeV). For situations in which contaminants emit beta particles of lower energy (e.g., facilities contaminated with nickel-63),the licensee will specifically determine detector efficiencies for those contaminants.The licensee will account for the effects of surface conditions on surface activity measurementsthrough the use of a source efficiency factor, in accordance with the guidance in ISO-7503-1, "Evaluation of Surface Contamination," Part 1, "Beta Emitters and Alpha Emitters," issued August 1998 (Ref. 13), and NUREG/CR-1507 (Ref. 10). The licensee general considers defaultsource efficiency factors of 0.5 for beta emitters greater than 0.4 MeV maximum energy and 0.25 for beta emitters between 0.150 MeV and 0.400 MeV maximum to be applicable to anticipated FNR contaminants and surface conditions. However, if contaminants or conditions are not consistent with use of these default values, the licensee will determine specific sourceefficiency factors and document them in the FSS design.The licensee will estimate detection sensitivities using the guidance in the MARSSIM (Ref. 9) and NUREG/CR-1507 (Ref. 10). The licensee will choose instrumentation and surveytechniques with the objective of achieving detection sensitivities of 25 percent of the criteria for structure surfaces, for both scanning and direct measurement, to ensure identification of areas of elevated activity having a size and activity level that could adversely impact the average residual activity level for the survey units.The licensee will follow guidance from equipment manufacturers and the American NationalStandards Institute (ANSI) N323-1978, "American National Standard Radiation Protection Instrumentation Test and Calibration," issued 1978 (Ref. 14), for calibration methods, calibration interval, and operational and background quality control checks. The licensee willestablish procedures to implement this guidance and will perform instrument calibrations usingstandards traceable to NIST or an equivalent standards organization.3.3.5.3 Data Quality Objectives The licensee designed its stated DQOs to achieve a 95-percent confidence level that therelease criteria are met. The survey design will be based on both Type I () and Type II ()decision errors of 5 percent. The DP describes data quality indicators for precision, accuracy, representativeness, completeness, and comparability as follows:*Precision is determined by comparison of replicate values from field measurements andsample analyses; the objective is a relative percent difference of 20 percent or less at 50 percent of the release criteria.*Accuracy is the degree of agreement with the true or known value; the objective for thisparameter is +/-20 percent at 50 percent of the release criteria.*Representativeness and comparability do not have numeric values. Performance isassured through selection and proper implementation of sampling and measurement techniques.*Completeness refers to the portion of the data that meets acceptance criteria and is thusacceptable for statistical testing; the objective for this parameter is 90 percent.3.3.5.4 Classifications of Areas by Contamination Potential For FNR areas determined to be impacted areas per guidance in the MARSSIM, the licenseeadopted the following definitions that describe three classifications of areas, according to contamination potential.(1)Class 1 areas are impacted areas that, prior to remediation, are expected to haveconcentrations of residual radioactivity that exceed the guideline value.(2)Class 2 areas are impacted areas that, prior to remediation, are not expected to haveconcentrations of residual radioactivity that exceed the guideline value. (3)Class 3 areas are impacted areas that have a low probability of containing residual activity. Typically levels will not exceed 25 to 35 percent of the guideline value.The licensee used facility history, including the Historical Site Assessment, issued 2003 (Ref.15), and radiological monitoring conducted during characterization and remedial activities as the bases for classification. Once the licensee obtains approval for the FSS plan through a subsequent license amendment request to the NRC, the licensee may make changes to theclassification of an area as long as the classification is changed to one of higher contamination potential. The licensee will obtain a license amendment pursuant to 10 CFR 50.90, "Applicationfor Amendment of License or Construction Permit," if the change would decrease an area classification (i.e., impacted to unimpacted, Class 1 to Class 2, Class 2 to Class 3, or Class 1 to Class 3), as discussed in Section 4.0.3.3.5.5 Identification of Survey Units A survey unit is a portion of a facility with common contaminants and contamination potentialand contiguous surfaces or areas. The licensee will identify survey units following remediation,at the time of FSS design. Table 4-3 of the DP provides a listing of facility areas that are currently expected to be included in the FSS, the estimated surface areas, anticipatedcontamination potential classifications, and the projected number of survey units within each area. The licensee developed this listing based on the historical assessment, preliminary survey data obtained in November 2002, and the characterization survey performed in April 2003. The DP notes that the licensee will determine actual survey unit boundaries andclassifications at the time of FSS design, and survey unit classifications and surface areas may change as characterization and remedial activities proceed. If classifications and boundaries change, the licensee will redesign the FSS for the affected areas and reevaluate data asnecessary.3.3.5.6 Demonstrating Compliance The null hypothesis recommended for use in the MARSSIM and selected by the licensee isstated, "The residual radioactivity in the survey unit exceeds the release criterion." Rejection of the null hypothesis by the statistical test therefore concludes that the residual activity does notexceed guidelines and the survey unit satisfies requirements for unrestricted release.The licensee will use nonparametric statistical tests recommended in the MARSSIM todemonstrate that radiological conditions satisfy the established criteria. One of the tests is theWilcoxon Rank Sum (WRS) test. The licensee may use the WRS test when a specific radionuclide of concern is present in background at a concentration greater than 10 percent ofthe guideline level and when the measurement is not radionuclide specific (e.g., for direct measurements of total surface activity). The licensee may use the Sign test when the radionuclide of concern is not present in background at a significant fraction (i.e., less than 10 percent) of the guideline level. The Sign test will also be used when evaluating data basedon the Unity Rule and may be used for surface activity data representing multiple surface media. Both of these tests are applicable to the FNR facility FSS, and the licensee will not beable to evaluate FSS data using statistical tests without first obtaining NRC approval. Thelicensee will select a specific test method when designing the FSS. 3.3.5.7 Background Reference Areas and MaterialsThe licensee will determine background contributions if (1) the residual contamination includesa radionuclide that occurs in background or (2) measurements are not radionuclide specific.

The licensee anticipates that the FSS will require multiple reference areas and materials. Forapplications involving the WRS test, reference areas will be of the same material as the survey unit being evaluated, but without a history of potential contamination by licensed operations. The licensee will obtain a set of reference measurements for each instrument used for survey unit evaluation. For applications involving the Sign test, sufficient background determinationswill be made for each media or surface material and with each instrument to provide an averagebackground level that is accurate to within +/-20 percent (usually requires a minimum of 8 to 10 measurements). The licensee will identify reference area and background requirements at thetime of individual survey unit FSS design.3.3.5.8 Final Status Survey Design 3.3.5.8.1 Sample Size and Sampling Locations The licensee provided adequate information that will be used for determining the data needs forthe statistical tests for each survey unit. The licensee indicated that the FSS design for thatsurvey unit will document the following information:*calculation of the relative shift (/)*/ = DCGL - lower bound of gray region*DCGL, as the gross or nuclide-specific release criteria

  • lower bound of the gray region initially selected as half of the DCGL as recommended bythe MARSSIM
  • determined empirically from actual survey data; however, for planning purposes, equalsa value of 25 percent of the DCGL*decision errors established by DQOs for this project of 0.05 for both Type I and Type IIerrors*determination of the number of data points required as obtained from MARSSIM Tables5.3 (WRS test) and 5.5 (Sign test)The MARSSIM recommends a triangular measurement or sampling pattern to increase theprobability of identifying small areas of residual activity. The licensee will use this type oftriangular pattern for the FSS, except where dimensions and/or other factors related to a specific survey unit require use of an alternate pattern. If the systematic pattern does notprovide sufficient data points to satisfy the number determined as outlined above, the licenseewill locate additional data points using a random-number technique. 3.3.5.8.2 Scan SurveysLicensee data collected in support of the FSS of structure surfaces will consist of scans toidentify locations of residual contamination, direct measurements of beta surface activity, and measurements of removable beta surface activity. The FSS data collected by the licensee for open land (soil) areas will consist of scans to identify locations of residual contamination andsamples of soil, analyzed for potential contaminants. The licensee will obtain additionalmeasurements and samples as necessary to supplement the information from these typical survey activities.The licensee will use gas-flow proportional detectors for beta surface scans. Floor monitorswith 580-cm 2 detectors will be used for floor and other larger accessible horizontal surfaces;hand-held 125-cm 2 detectors will be used for surfaces not assessable with the floor monitor. When scanning, (1) the detector will be within 0.5 cm of the surface (if surface conditionsprevent this distance, the detection sensitivity for an alternate distance will be determi ned andthe scanning technique adjusted accordingly), (2) scanning speed will be no greater t han onedetector width per second, and (3) audible signals will be monitored and locations of elevateddirect levels identified for further investigation. The licensee committed to the minimum scan coverages of 100 percent for Class 1 surfaces, 25 percent for Class 2 surfaces, and 10 percent for Class 3 surfaces. Coverage for Class 2 and Class 3 surfaces will be biased towards areasconsidered by professional judgment to have the highest potential for contamination.The licensee will use NaI gamma scintillation detectors (2 inch x 2 inch) for gamma surfacescans of structures and open land areas to identify locations of residual surface activity. When scanning, (1) the detector will be moved in a serpentine pattern, while advancing at a rate ofapproximately 0.5 meters per second, (2) the distance between the detector and the surface willbe maintained within 5 centimeters of the surface, and (3) audible signals will be monitored andlocations of elevated direct levels identified for further investigation. The licensee committed to the minimum scan coverages of 100 percent for Class 1 surfaces, 25 percent for Class 2 surfaces, and 10 percent for Class 3 surfaces. Coverage for Class 2 and Class 3 surfaces willbe biased toward areas considered by professional judgment to have the highest potential for contamination.3.3.5.8.3 Direct Measurements and Sampling The licensee will perform direct measurement of beta surface activity at designated locationsusing a 125-cm 2 gas-flow detector. Measurements will be conducted by integrating the countover a 1-minute period. Where adverse surface conditions may result in underestimating activity by direct measurements, the licensee will obtain surface samples for laboratory analyses. Thelicensee's FSS design will identify the need for such sampling for specific survey units.The licensee will collect a smear sample for removable activity at each direct surface activitymeasurement location with a 2-inch diameter cloth or paper filter by wiping a 100-cm 2 surfacearea using moderate pressure. Dampened smears will be used to sample for removable tritiumactivity.The licensee will obtain samples of surface (upper 15 centimeters) soil from selected locationsusing a hand trowel or bucket auger. Approximately 500 to 1000 grams of soil will be collectedat each sampling location. 3.3.5.9 Data AssessmentThe licensee committed to review radiological data needed to support the FSS to assure that thetype, quantity, and quality are consistent with the survey plan and design assumptions.

Data standard deviations will be compared with the assumptions made in establishing thenumber of data points.The licensee will compare individual and average data values with guideline values and confirmproper survey area classifications.The licensee will investigate individual measurement data in excess of the guideline level forClass 2 areas and in excess of 25 percent of the guideline for Class 3 areas. Anomalies and deviations from design assumption and plan requirements will be identified. Need forinvestigation, reclassification, remediation, and/or resurvey will be determined. In addition, thelicensee will initiate a corrective action will be initiated, as appropriate, and repeat the dataconversion and assessment process for new data sets.3.3.5.10 Final Status Survey Report The licensee will prepare an FSS report describing the survey procedures and findings forsubmission to the NRC in support of license termination. The FSS report will provide a completerecord of the facility's radiological status and a comparison to the site release criteria. Thelicensee's FSS report will provide a summary of any ALARA analysis, survey data results, andoverall conclusions, which collectively demonstrate that the FNR facility meets the radiological criteria for unrestricted use. The FSS report will include information such as the number andtype of measurements, basic statistical quantities, and statistical test results. It will also containadditional detail to enable an independent or third party re-creation and evaluation of the survey results and a determination as to whether the site release criteria have been met.The following outline from Section 4.15 of the DP illustrates a general format that the licenseemay use for the FSS report:*a summary of the results of the FSS

  • a discussion of any changes in the FSS process that were proposed in the licensetermination plans or other prior submittals*a description of the method used to determine the number of samples for each survey unit
  • a summary of the assumed parameters used to calculate the number of samples and ajustification for these valuesThe FSS report will also provide the results for each survey unit, including the following:*the number of survey samples collected for the survey unit*a map or drawing of the survey unit showing the reference system and random startsystematic sample locations for Class 1 and 2 survey units, and random locations shown forClass 3 survey units and reference areas *measured sample concentrations*statistical evaluation of the measured concentrations
  • judgmental and miscellaneous sample data sets reported separately from those samplescollected for performing the statistical evaluation*discussion of anomalous data including any areas of elevated direct radiation detectedduring scanning that exceeded the investigation level or measurement locations in excess of the DCGL*a statement that a given survey unit satisfied the DCGL and the elevated measurementcomparison if any sample points exceeded the DCGL*a description of any changes in initial survey unit assumptions relative to the extent ofresidual radioactivity*a description of the investigation conducted when the data from a survey unit fail toascertain the reason for the failure and a discussion of the impact that the failure has on the conclusion that the facility was ready for final radiological surveys*a description of the impact a survey unit failure has on other survey unit information and thereason for the failureThe licensee may adjust this outline to more clearly present the information. The level of detailwill be sufficient to clearly describe the FSS program and certify the results.3.3.5.11 Change Control The staff reviewed the DP to determine whether it listed sufficient criteria to establish the typesof changes to equipment, structures, system components, and procedures that would bepermissible without prior NRC approval. The staff, recognizing that change control criterianeeded to support reactor operations were not well suited for determining the types of changes expected to occur during decommissioning, issued Amendment No. 47 for the FNR on January 29, 2004, to maintain the authority to make changes to the facility and procedures without priorCommission approval as contained in 10 CFR 50.59. Therefore, the following change controlcriteria that would support changes that may be needed to implement the FSS during decommissioning would not require prior NRC approval:*The licensee may make changes to the DP without prior approval provided the proposedchanges do not--require Commission approval pursuant to 10 CFR 50.59

-use a statistical test other than the Sign test or WRS test for evaluation of the FSS

-increase the radioactivity level, relative to the applicable DCGL, at which aninvestigation occurs -reduce the coverage requirements for scan measurements-decrease an area classification (i.e., impacted to unimpacted, Class 1 to Class 2,Class 2 to Class 3, or Class 1 to Class 3)-increase the Type I decision error

-result in more than a minimal increase in the environmental consequences notpreviously evaluated in the final SAR (as updated)-foreclose the release of the site for possible unrestricted use*The licensee shall submit reports of any characterization surveys performed that were notpart of the license amendment application and shall submit the completed FSS plan for review prior to performing the FSS.The staff finds that the change control criteria proposed by the licensee will adequately facilitatechanges needed to implement the FSS in a manner that ensures both the safety of workers and the public and facilitates timely decommissioning of the FNR.3.3.5.12 Conclusions The staff has reviewed the licensee's DP concerning the planning of the FSS. The staff findsthat the licensee has adequate experience to develop and implement an acceptable MARSSIMFSS. Once the licensee develops the FSS plan, it will present the plan for review and approvalprior to implementation. The NRC staff concludes this aspect of the DP meets the requirementsof 10 CFR 50.82(b)(4)(iii) and is therefore acceptable.3.4 Estimated Cost The licensee stated that decommissioning of the FNR will be accomplished withoutdismantlement of the building. Table 1-1 of the DP presents the detailed estimated cost todecommission the FNR licensed areas. The factors used in these cost estimates were based upon a detailed cost estimate. Using the "High" cost in Table 1-1 of the DP, the licensee estimated that the project will cost up to $9,781,173. Based on the given "High" estimate, theDP states that UM is committed to providing funding for decommissioning of the FNR, in accordance with 10 CFR 50.75(e)(iv).3.4.1 Conclusions The staff has reviewed the licensee's decommissioning cost estimate and finds that the costestimates are consistent with the scope of work covering decommissioning of the FNR. The licensee stated that the UM Regents have specifically approved the expenditure of funds frominvestment proceeds sufficient to cover the "High" cost estimate. The staff concludes that UM is committed to providing acceptable funding for decommissioning of the FNR.3.5 Quality Assurance 3.5.1 Overview Section 1.3.4.1 of the DP briefly describes the quality assurance programs used duringdecommissioning, summarized as follows:*A quality assurance program is applied to the design, fabrication, construction, and testing ofstructures, systems, and components of the facility. These quality assurance requirementswould apply to the remediation activities conducted. *A quality assurance program, which may or may not be the same as the above-mentionedprogram, is applied to the design, purchase, fabrication, handling, shipping, storing, cleaning,assembly, inspection, testing operations, maintenance, repair, and modification of components of packaging used in the transportation of licensed material.*Additional quality assurance requirements are applied to the FSS and associateddocumentation (e.g., characterization information used in the design of the FSS) to ensure that data and the analysis of the data provided to the NRC in the FSS report are accurateand complete. 3.5.2 Quality Assurance for Design, Construction, Testing, Modification, and MaintenanceThe FNR has a quality assurance program, as discussed in Section 1.3.4.2 of the DP, thatmeets the requirement in 10 CFR 50.34, "Contents of Applications; Technical Information," forestablishing and executing a quality assurance program for the design, construction, testing, modification, and maintenance of a research reactor. The descriptions of the managerial and administrative controls will result in a revision to the current quality assurance program. TheFNR will continue to maintain this quality assurance program for the design, construction,testing, modification, and maintenance (including remediation activities) of the reactor.UM will continue to require that all contractors and subcontractors participating in design,construction, testing, modification, and maintenance (including remediation) activities follow the established quality assurance program. Contractors and subcontractors may recommend or request changes to the quality assurance program. UM may or may not make changes to the quality assurance program after review against applicable guidance or standards recommended.

Changes to the quality assurance program will be approved as discussed in Section 2.4 of theDP.3.5.3Quality Assurance for Packaging, Preparation for Shipment, and Transportation ofLicensed MaterialSubpart H, "Quality Assurance," of 10 CFR Part 71 specifies the requirements for packaging,preparation for shipment, and transportation of licensed material. The managerial and administrative controls the FNR has established to satisfy the requirements of this subpart, described in Section 2.4 of the DP, differ slightly from those previously used. The NRC hasapproved the current FNR quality assurance program as required by 10 CFR 71.101(c). The licensee will follow the existing quality assurance program and maintain it through timelyrenewal, as necessary, to support packaging, preparation for shipment, and transportation of licensed material during remediation activities.UM will continue to require that all contractors and subcontractors participating in packaging,preparation for shipment, and transportation of licensed material follow the approved quality assurance program. Contractors and subcontractors may recommend or request changes tothe quality assurance program. UM may or may not make changes to the quality assurance program after review against the requirements of 10 CFR Part 71, Subpart H. The licensee willsubmit revisions to the quality assurance program to the NRC for approval as required by10 CFR 71.101(c) prior to implementation and use for the packaging, preparation for shipment,and transportation of licensed materials.UM may elect to use a contractor's or subcontractor's quality assurance program to fulfill therequirements contained in 10 CFR Part 71, Subpart H, after verification that the contractor's orsubcontractor's quality assurance program is acceptable to UM and has been approved by the

NRC.3.5.4 Quality Assurance for Final Status Survey and Associated Documentation3.5.4.1 General UM is responsible for developing a FSS quality assurance program and associateddocumentation (e.g., characterization information used in the design of the FSS). This program will be reviewed and approved as described in Section 2.4 of the DP. The FSS qualityassurance program will incorporate the appropriate regulatory requirements applicable to theplanning and conduct of radiological surveys necessary for the termination of the FNR license and the release of the site for unrestricted use. The quality assurance program implements the appropriate criteria in Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50, "Domestic Licensing of Production and UtilizationFacilities." The following sections describe the required components of the FSS qualityassurance program.3.5.4.2 Organization Section 2.4 of the DP identifies the written definitions of authorities, duties, and responsibilities ofmanagerial, operations, and safety personnel; a defined organizational structure; assigned responsibility for review and approval of plans, specifications, designs, procedures, data, andreports; and assigned responsibility for procurement and oversight of services (e.g., analyticallaboratory). The licensee will give personnel assigned organizational responsibility forperforming quality assurance functions the necessary independence and authority to allow them to identify quality problems; to initiate, recommend, and provide solutions; and to verify implementation of solutions.The reactor manager has the authority and responsibility for implementing all aspects of the FSSquality assurance program. The reactor manager will ensure that survey activities meet therequirements outlined in the FSS quality assurance program to safeguard the decommissioning staff, the UM community, and the public. The reactor manager will regularly review theadequacy of the FSS quality assurance program and provide an assessment to the Director and the review committee. The reactor manager will inform the appropriate UM decommissioningstaff and contractors of decommissioning activities related to the FSS quality assurance program. The project manager will ensure that the contractor complies with the FSS quality assuranceprogram and satisfies the objectives and requirements for FSS. Furthermore, the project manager is responsible for ensuring that all activities are performed in a manner to permit thetermination of the FNR license and the release of the site for unrestricted use. In accordance with the American Society of Mechanical Engineers "Quality Assurance Requirements for Nuclear Facility Applications," issued 2001 (Ref. 16), the individual(s) or organization(s)responsible for establishing and executing the FSS quality assurance program may delegate any or all of the work to others but will otherwise retain responsibility.3.5.4.3 Written Quality Assurance Program The licensee will establish a documented quality assurance program for the FSS and associateddocumentation (e.g., characterization information used in the design of the FSS) at the earliest practical time, consistent with the schedule for accomplishing the activities. The licensee willdocument this quality assurance program through written polices, procedures, or instructions

and will execute it through the conduct of FSS activities and creation of associateddocumentation in accordance with those policies, procedures, or instructions. Activities for the FSS and creation of associated documentation affecting quality will be accomplis hed undersuitably controlled conditions. Controlled conditions included the use of appropriate equipment, suitable environmental conditions for accomplishing the activity, and assurance that prerequisites for the given activity have been satisfied. The quality assurance program willprovide for any special controls, processes, survey equipment, tools, and skills to attain therequired quality of activities and items and for verification of that quality.3.5.4.4 Training Personnel will be qualified for their assigned duties before working independently or will beunder the direct supervision of a qualified individual. Personnel performing special processes will be qualified according to specific codes and standards or in accordance with nationalconsensus documents. Qualification will include proficiency demonstrated by each individual,both initially and then periodically. Qualification will also be demonstrated when required by thedesignated codes or standards.The licensee will maintain training records that include the trainee's name, dates of training,types of training, test results, protective equipment use authorizations, and instructors' names.

Care will be taken to ensure that properly qualified instructors conduct all training. As theprimary criterion, persons responsible for the presentation of training should have knowledge and experience in the process or subject matter. It is desirable that trainers also have the presentation skills or classroom conduct appropriate to the level of the training being presented. For those with limited background in training, early instruction should be monitored and feedback should be provided.3.5.4.5 Quality Assurance Records The licensee will ensure that sufficient records are specified, prepared, reviewed, authenticated,and maintained to reflect the achievement of the required quality. Records will includedocuments such as operating logs, results of reviews, inspections, tests, assessments, work performance monitoring, and material or sample analyses. Records will be identifiable,available, and retrievable. The records will be reviewed to ensure their completeness and abilityto serve their intended function. Requirements will be established concerning record collection,safekeeping, retention, maintenance, updating, location, storage, preservation, administration, and assigned responsibility. Requirements will be consistent with applicable regulations, as wellas the potential for impact on quality and radiation exposure to workers and the public.The licensee will identify documents that require control, including policies, procedures, orinstructions that specify quality requirements or describe activities affecting quality, such as instructions, procedures, and drawings. Qualified personnel will review policies, procedures, orinstructions (including revisions) for conformance with technical requirements and quality systemrequirements and approve them as discussed in Section 2.4 of the DP. The personnel performing relevant activities will ensure the currency of policies, procedures, or instructionsrequiring control. The licensee will take measures to ensure that personnel understand thedocument controls to be used. Obsolete or superseded documents will be identified andmeasures will be taken to prevent their use.The licensee will control all documents related to the FSS using appropriate policies,procedures, or instructions. All significant changes to such documents will be similarlycontrolled. This documentation normally would include a survey plan, survey packages, survey results, and a survey report.3.5.4.6 Control of Measuring Equipment Measures will be established to assure that instruments and other measuring devices used inactivities affecting quality are properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits. The licensee will base the selection of instruments on the type of radiation emitted for theradionuclides of interest, as well as the required range, accuracy, and tolerance needed to demonstrate conformance to specified requirements. Selection and use of instrumentation for the FSS will also be based upon the need to ensure that the residual radioactivity remaining onsite meets the release criteria. Table 3-7 lists the instrumentation intended for use for the FSSand associated documentation (e.g., characterization information used in the design of the final survey), along with estimated detection sensitivities. Other instruments, which are the functional equivalent of those listed, will also be acceptable.Calibration procedures will identify or reference required accuracy. Methods of evaluating theaccuracy of instrumentation will be defined in procedures and will follow ANSI N323-1978 (11). The calibration method and interval of calibration for instruments will be defined, based on thetype of equipment, stability characteristics, required accuracy, intended use, manufacturer'srecommendations, and other conditions affecting capability, and will follow ANSI N323-1978. Out-of-calibration and defective instruments will be removed from service and not used until theyhave been repaired and recalibrated. The licensee will repair or replace any instrumentsconsistently found to be out of calibration.Measuring instruments will be calibrated at prescribed time periods or immediately before useand whenever the accuracy of the equipment is suspect. Calibration will be performed usingstandards traceable to NIST or an equivalent standard organization. Instruments found to be out of calibration will require a documented evaluation, commensurate with the significance ofthe condition, of the validity of data obtained with that instrument since its previous acceptableperformance. Instruments will be properly handled and stored to maintain accuracy according toANSI N323-1978. The licensee will suitably mark or otherwise identify instruments to indicate calibration status.Operational and backgr ound c hecks will be performed at the beginning of each day of FSSactivity and whenever there is reason to question instrument performance. T hese che cks shouldfollow ANSI N323-1978.3.5.4.7 Audits and Corrective Actions Project audits will be planned and conducted using criteria that describe acceptable workpractices, including performance. Audits will verify compliance with applicable requirements ofthe FSS quality assurance program and will determine its effectiveness. The scheduling ofaudits and allocation of resources will be based on the work status, risk, and complexity of theitem or process being assessed. Audits will be performed and results reported as described inSection 2.4 of the DP. Conditions adverse to quality will be identified to the reactor managerpromptly and corrected as soon as practicable. Significant conditions adverse to quality will beidentified to the licensee's review committee as soon as practicable, along with the cause of the condition, when known, and corrective actions taken to prevent recurrence.3.5.4.8 Conclusions The currently approved quality assurance program will be in place for activities leading up to theFSS, such as remediation and transportation of licensed material. The staff has reasonable assurance that an adequate quality assurance plan is in place and implemented in accordance with 10 CFR 50.82(b)(4)(v) for these activities.The information presented in the DP provides reasonable assurance to the staff that a FSSquality assurance program constructed according to the stated requirements will adequatelyaddress the necessary quality functions associated with decommissioning activities in accordance with 10 CFR 50.82(b)(4)(v). 3.6 Physical Security The regulations in 10 CFR 73.67(c)(1) require facilities to maintain a physical security plan whenthey possess special nuclear materials of moderate strategic significance or 10 kilograms or more of special nuclear material of low strategic significance. Because all special nuclear material in the form of reactor fuel covered by the license for the FNR has been removed, and the license has been amended for no possession of reactor fuel (Amendment No. 47), a physical security plan is not required.It is recognized that the regulations in 10 CFR Part 20, Subpart I, "Storage and Control ofLicensed Material," apply to the remaining byproduct and special nuclear materials possessed by the FNR. All FNR licensed materials that are in storage will be secured from unauthorizedaccess or removal, and licensed materials that are not in storage will be under the control andconstant surveillance of authorized FNR personnel as required by 10 CFR Part 20.3.6.1 Conclusions Based on the NRC staff's review, the licensee has acceptable security access controls toprevent inadvertent exposure to workers and members of the public. 4.0 ADDITIONAL LICENSE CONDITIONSThe regulations in 10 CFR 50.82(b)(5) state in part that the licensee's DP will be approved bylicense amendment subject to such conditions and limitations as the NRC deems appropriateand necessary. Based on the requirements of the regulations and the staff's review of the licensee's application, the staff has added the following conditions to the UM FNR license:The license is amended to approve the decommissioning plan described in the licensee'sapplication dated June 23, 2004, as supplemented on January 05, 2006, and authorizesinclusion of the decommissioning plan as a supplement to the Safety Analysis Report pursuant to 10 CFR 50.82(b)(5).A license amendment pursuant to 10 CFR 50.59 shall be obtained for changes to thisdecommissioning plan if the change would:*Require Commission approval pursuant to 10 CFR 50.59;

  • Use a statistical test other than the Sign test or Wilcoxon Rank Sum test for evaluationof the final status survey;*Increase the radioactivity level, relative to the applicable derived concentrationguideline level, at which an investigation occurs;*Reduce the coverage requirements for scan measurements;
  • Decrease an area classification (i.e., impacted to unimpacted, Class 1 to Class 2,Class 2 to Class 3, or Class 1 to Class 3);*Increase the Type I decision error;
  • Result in more than a minimal increase in the environmental consequences notpreviously evaluated in the final safety analysis report (as updated);*Foreclose the release of the site for possible unrestricted use.

The licensee shall submit reports of any characterization surveys performed that are notpart of the license amendment application and shall submit the completed final statussurvey plan for review prior to performing the final status survey.The above license conditions make the licensee's DP part of the Safety Analysis Report for thefacility in accordance with the regulations, help to ensure that changes to the DP that mayimpact compliance with the release criteria in the regulations in Part 20 are not made without NRC review, and ensure that important information to the decommissioning process st ill underdevelopment by the licensee are submitted to the NRC when complete. 4.1 ConclusionsThe staff has added requirements to the UM FNR license in accordance with the regulations in10 CFR 50.82(b)(5). The staff concludes that these license conditions are necessary to meet the requirements of 10 CFR 50.82(b)(5) and to allow the licensee to develop the final radiological survey and documentation necessary to permit the staff to make the required findings to terminate the license in accordance with 10 CFR 50.82(b)(6).5.0 TECHNICAL SPECIFICATIONS The licensee's organization for decommissioning is changing substantially. To support thesechanges, the licensee proposed revisions to TS 6.0, "Administrative Controls." Those changes were issued with Amendment No. 49 to Facility License No. R-28 (Ref. 17). The NRC will issueAmendment No. 49 concurrently with the decommissioning amendment approving the UM FNR DP. 5.1 Conclusions With the issuance of Amendment No. 49 to Facility License No. R-28 for the UM FNR reactor,appropriate changes have been made to support the UM FNR DP and the safe decommissioning of the reactor.

6.0 ENVIRONMENTAL CONSIDERATION

The Commission has prepared an EA and Finding of No Significant Impact, published in the FRon February 6, 2006 (71 FR 6104-6105). On the basis of the EA and this safety evaluation, the Commission has determined that no environmental impact statement is required and thatissuance of this license amendment approving decommissioning will have no significant adverseeffect on the quality of the human environment.

7.0 CONCLUSION

SBased on the staff's review of the licensee's application for approval of decommissioning, thestaff finds that the licensee is adequately cognizant of its continuing responsibilities to protectthe health and safety of both workers and the public from undue radiological risk. The DP provides reasonable evidence that the licensee is prepared to dismantle the reactor and dispose of all significant reactor-related radioactive materials in accordance with applicable regulationsand applicable NRC guidance.The staff concludes that the choice of the DECON decommissioning alternative is acceptableand meets the requirements of 10 CFR 50.82(b)(4)(i) for decommissioning without significant delay.The staff concludes that the DP provides acceptable organizational structure and control todecontaminate and dismantle the FNR while maintaining due regard for protecting the public,environment, and workers from significant radiological risk. Furthermore, the staff concludes that the licensee's plan for radiation protection and radioactive material and waste management is acceptable based on the use of standard guidance and practices for such programs. The staff finds the personnel training program that FNR proposed in the DP to be acceptablebecause its scope covers all aspects of decommissioning activities that need to be performed safely. The industrial safety program and procedural and equipment controls are consistent with such programs at decommissioning reactors and are therefore acceptable. The staff concludes that potential radiological consequences attributable to the types of accidents that could occurduring decommissioning are well within acceptable limits. The staff concludes that the licensee's DP contains a description of the controls and limits on procedures and equipment to protect occupational and public health and safety as required by 10 CFR 50.82(b)(4)(ii).The staff concludes that the licensee has adequately described the radiological status of theFNR facility and has proposed acceptable release criteria for the FNR facility. The licensee hasacceptably described t he tasks, sequence of activities, and schedule needed to decommissionthe FNR facility. The staff also concludes that the licensee has provided an acceptabledescription of its planned final radiation survey as required by 10 CFR 50.82(b)(4)(iii).The staff concludes that the licensee has provided, in accordance with 10 CFR 50.82(b)(4)(iv),an acceptable updated cost estimate for the DECON decommissioning alternative and has an acceptable plan for assuring the availability of adequate funds for the completion ofdecommissioning.The licensee has provided a description of TSs, quality assurance provisions, and physicalsecurity plan provisions to be in place during decommissioning. The staff has determined that these aspects of the DP meet the regulations in 10 CFR 50.82(b)(4)(v). Therefore, based on thediscussion above, the staff concludes that the licensee's DP meets the requirements of10 CFR 50.82 (b)(4).The staff has concluded, on the basis of the considerations discussed above, that (1) becausethe amendment does not involve a significant increase in the probability or consequences ofaccidents previously evaluated, or create the possibility of a new or different kind of acci dentfrom any accident previously evaluated, and does not involve a significant reduction in a margin of safety, the amendment does not involve a significant hazards consideration, (2) there isreasonable assurance that the health and safety of the public will not be endangered by theproposed activities, and (3) such activities will be conducted in compliance with theCommission's regulations and the issuance of this amendment will not be inimical to thecommon defense and security or the health and safety of the public. ABBREVIATIONSAECairborne effluent concentrationALARA as low as reasonably achievable ANSI American National Standards Institute APactivation product Bqbecquerel

CFR Code of Federal RegulationsCicurie(s) cmcentimeter(s)

cm 2square centimeter(s)Cscesium DAWdry active waste DCGL derived concentration guideline level DECON decontamination decommissioning option DOEU.S. Department of Energy DP decommissioning plan dpmdisintegration(s) per minute DQOdata quality objective FNRFord Nuclear Reactor FPfission product

FRFederal RegisterFSSfinal status survey

ft 3cubic foot/feetggram(s) hhour(s)

HEPA high-efficiency particulate air HIChigh-integrity containers HSSIheavy section steel irradiation HVACheating, ventilation, and air conditioning in.inch(es)

JHAjob hazard analysis kgkilogram(s)

Lliter(s) lbpound(s)

LSAlow specific activity

m 2square meter(s)MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual mCimillicurie(s)

MeVmegaelectron Volts MIOSHAMichigan Occupational Safety and Health Act of 1974 mLmilliliter mremmillirem mSvmillisievert Mwt megawatt thermal NaIsodium iodide NIST National Institute of Standards and Technology NRC U.S. Nuclear Regulatory Commission pCipicocurie(s)PMLPhoenix Memorial Laboratory PPEpersonal protective equipment Rroentgen RIFLSreactor irradiation facility for large samples RSO radiation safety officer RWP radiation work permit SARsafety analysis report SSstainless steel TEDE total effective dose equivalent (see 10 CFR Part 20)

TS technical specification UMUniversity of Michigan WRSWilcoxon Rank Sum yryear REFERENCES1.University of Michigan, Ford Nuclear Reactor-Amendment, Decommissioning Plan,June 18, 2004.2.University of Michigan, Decommissioning Plan for the Ford Nuclear Reactor, Rev. 00,June 23, 2004.3.University of Michigan, Decommissioning Plan for the Ford Nuclear Reactor, Rev. 01,January 5, 2006.4.University of Michigan, Additional Detail on the In Situ Gamma Spectroscopy Data,January 10, 2006.5.U.S. Nuclear Regulatory Commission, NUREG-1700, Rev. 1, Standard Review Plan forEvaluating Nuclear Power License Termination Plans, Appendix 2, Washington DC, April 2003. 6.U.S. Nuclear Regulatory Commission, NUREG-1537, Rev. 0, Guidelines for Preparingand Reviewing Applications for the Licensing of Non-Power Reactors, Appendix 17.1, Washington, DC, February 1996.7.U.S. Nuclear Regulatory Commission, Circular No. 81-07, Control of RadioactivelyContaminated Material, Washington, DC, May 14, 1981.8.U.S. Nuclear Regulatory Commission, Information Notice No. 85-92, Surveys ofWastes Before Disposal from Nuclear Reactor Facilities, Washington, DC, December

2, 1985.9.U.S. Nuclear Regulatory Commission, NUREG-1575, Rev. 1, Multi-Agency RadiationSurvey and Site Investigation Manual (MARSSIM), Washington, DC, August 2000.10.U.S. Nuclear Regulatory Commission, NUREG/CR-1507, Final, Minimum DetectableConcentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, Washington, DC, June 1997.11.U.S. Nuclear Regulatory Commission, NUREG-0586, Final Generic EnvironmentalImpact Statement on Decommissioning of Nuclear Facilities, Washington, DC, 1988.12.U.S. Nuclear Regulatory Commission, Regulatory Guide 8.15, Rev. 1, AcceptablePrograms for Respiratory Protection, Washington, DC, October 1999.13.International Organization for Standardization, Evaluation of SurfaceContamination-Part 1: Beta Emitters and Alpha Emitters (First Edition), ISO-7503-1, August 1998.14.American National Standards Institute, ANSI N323-1978, American National Standard Radiation Protection Instrumentation Test and Calibration, 1978.15.CH2M HILL, Inc., Historical Site Assessment, Ford Nuclear Reactor, North Campus,University of Michigan, Final Report, Richland, Washington, 2003.16.American Society of Mechanical Engineers, Quality Assurance Requirements forNuclear Facility Applications, New York, NY, 2001.17.U.S. Nuclear Regulatory Commission, Amendment No. 49 to Facility Operating LicenseNo. R-28 University of Michigan Ford Nuclear Reactor, Appendix A, May , 2006.

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