ML042740512

Request for License Termination for the Alan J. Blotcky Reactor Facility

ML042740512
Person / Time
Site:	05000131
Issue date:	09/21/2004
From:	Washko A US Dept of Veterans Affairs, Nebraska-Western Iowa Health Care System
To:	Document Control Desk, NRC/FSME
References
636/151
Download: ML042740512 (159)

Text

Omaha 4101 Woolworth Avenue DEPARTMENT OF VETERANS AFFAIRS Omaha NE 681051873 NEBRASKA-WESTERN IOWA HEALTH CARE SYSTEM Lincoln 600 S 70th Street Lincoln NE 68510-2493 Grand Island 2201 N Broadwell Avenue Grand Island NE 68803-2196 In Reply Refer To: 636/151 September 21, 2004 U.S. Nuclear Regulatory Commission Attention: Document Control Desk (03H8) 11555 Rockville Pike Rockville, MD 20852 RE: Request for License Termination for the Alan J. Blotcky Reactor Facility, License R-57, Docket #50-131

1. We are requesting the NRC license termination of the Alan J. Blotcky Reactor Facility (License R-57, Docket #50-131 after the appropriate approval of the decommissioning plan and the decommissioning process.

2. Attached is the decommissioning and decontamination plan that has been developed for and approved by the Reactor Safety Committee of the Alan J.

Blotcky Reactor Facility.

3. We trust this communication is in compliance with all appropriate US Nuclear Regulatory Commission regulation specifications. If not, please respond as appropriate.

eG/

Al Washko Director Cc: Alexander Adams, Jr.

.4D20

Department of Veterans Affairs Decommissioning and Decontamination Plan for the Alan J. Blotcky Reactor Facility March 2004

Table of Contents Index of Tables 4

Index of Figures 5

List of Abbreviations 6

Index of Tables 4

1 Executive Summary 9

1.1 Licensee/Site Name and Address 9

1.2 Site Description 9

1.3 Summary of Licensed Activities 9

1.4 Nature ofSources 9

1.5 Location and Storage of Licensed Radionuclides 10 1.6 Decommissioning Objective 10 1.7 Derived Concentration Guideline Levels (DCGLs) 10 1.8 ALARA Evaluations 10 1.9 Decommissioning Timeline 11 1.10 License Amendmentfor Decommissioning 11 2

Facility Operating History 12 2.1 License Status 12 2.2 License History 12 2.3 Previous Decommissioning Activities 16 2.4 Spills and Burials 16 3

Facility Description 17 3.1 Site Location 17 3.2 Population Distribution 27 3.3 Current and Future Land Use 28 3.4 Meteorology And Climatology 28 3.5 Geology 32 3.6 Surface Water Hydrology 33 3.7 Groundwater Hydrology 33 3.8 Natural Resources 35 4

Current Radiological Status of the Facility 36 l

4.1 Contaminated Structures 40 4.2 Contaminated Systems and Equipment 43 4.3 Reactor Pool and Components 44 4.4 Surface Soil Contamination 45 4.5 Subsurface Soil Contamination 45 4.6 Surface Water 47 4.7 Groundwater 47 5

Dose Modeling Evaluations 48 5.1 Unrestricted Release Using Screening Criteria 48 5.2 Unrestricted Release Using Site-specific Information 49 5.3 Restricted Release Using Site-specific Information 49 5.4 Release Involving Alternative Criteria 49 6 Alternatives Considered and Rationale Supporting the Decommission Strategy Selected 50 6.1 Decommissioning Objective and Strategy Selected 50 6.2 Alternatives Considered 50 6.3 Rationalefor Chosen Alternative 51 7 ALARA Analysis 52 7.1 Description of How the Licensee Will Achieve a Decommissioning Goal Below the Dose Limit 52 7.2 Cost Benefit Analysis 53 8 Planned Decommissioning Activities 54 8.1 Contaminated Structures 66 8.2 Contaminated Systems 70 8.3 Soil and Exterior Areas 72 8.4 Surfaces and Groundwater 73 8.5 Schedule 73 9

Project Management and Organization 74 9.1 Decommissioning Management Organization 74 9.2 Decommissioning Task Management 84 9.3 Decommissioning Management and Oversight Positions Qualifications 85 9.4 Training 88 9.5 Contractor Support 91 10 Radiation Safety and Health Program 94 10.1 RadiationSafetyProgram 94 2

10.2 Nuclear Criticality Safety 114 10.3 Health PhysicsAudits, Inspections, and Record Keeping 114 10.4 General Industrial SafetyProgram 116 11 Environmental Monitoring and Control Program 118 11.1 Environmental As Low As Reasonably Achievable (ALARA) Radiation Exposures Evaluation Program 119 11.2 EffluentMonitoringProgram 120 11.3 Effluent Control Program 122 12 Waste Management Program 125 12.1 General Waste Management Program 125 12.2 Radioactive Waste Management 125 12.3 Non-radioactive Hazardous Waste Management 135 12.4 Potential Environmental Response 139 12.5 Spent Fuel Management 139 13 QualityAssurance Program 140 13.1 The Decommissioning and Decontamination Project QA Organization 141 13.2 The Omaha VAMC Quality Assurance Program 142 13.3 Document Control 142 13.4 Control of Measuring and Test Equipment (M&TE) 143 13.5 Corrective Action 143 13.6 Records Management 144 13.7 Audits and Surveillances 144 14 Facility Radiation Surveys 146 14.1 Release Criteria 146 14.2 Site Characterization Reports 147 14.3 Remediation Action Support Surveys 149 14.4 Final Status Survey Design 150 14.5 Conduct of the Final Status Survey 152 14.6 Final Status Survey Report 153 15 Financial Assurance 154 16 Restricted Use/Alternate Criteria 155 17 References 156 17.1 Code of Federal Regulation (CFR) Parts 156 17.2 NUREGs 156 17.3 OtherReferenced Materials 156 3

Index of Tables Table 2.1, Authorized Radionuclides 14 Table 3.1, Functions of Areas/Rooms 18 Table 3.2, Included rooms outside of the AJBRF 18 Table 3.3, Douglas County, Nebraska 2000 Census Data 27 Table 3.4, Tornado Frequency Data 30 Table 4.1, Potential Radionuclides of Concern at the AJBRF 36 Table 4.2, Resin Sample Analysis Results Summary 39 Table 4.3, Summary of Background Levels 41 Table 4.4, Locations and Amounts of Contamination Levels 42 Table 4.5, Types and Activity of Radioactive Material Contamination 43 Table 8.1, Expected radiological and industrial safety hazards and mitigating measures 62 Table 9.1, Positions and Qualifications 86 Table 10.1, AJBRF Radiation Protection Instrumentation 97 Table 12.1, AJBRF Waste Disposition 127 Table 15.1, Decommissioning Project Cost Estimate 154 4

Index of Figures Figure 2.1, Reactor Facility Floor Plan 15 Figure 3.1a, USGS topographic map of Omaha, NE 22 Figure 3.1b, USGS topographic map of Omaha, NE (detail) 23 Figure 3.2, Site Plan of the Omaha VAMC 24 Figure 3.3, Omaha VAMC Roof Exhaust Locations 25 Figure 3.4, Reactor and Pit 26 Figure 3.5, Omaha Historic Tornado Touchdown Map 31 Figure 3.6, Test wells, relative elevations of the groundwater, and resulting hydraulic gradient.

34 Figure 4.1, Reactor Dose Rates 44 Figure 4.2, Soil Boring Locations 46 Figure 8.1 Preliminary Decommissioning Project Schedule (Gantt chart) 55 Figure 9.1, Decommissioning Organization 75 5

List of Abbreviations ACGIH American Conference of Government and Industrial Hygienists ACM Asbestos Containing Material ACOS/R Associate Chief of Staff for Research AD Alarming Dosimeter AJB Alan J. Blotcky AJBRF Alan J. Blotcky Reactor Facility ALARA As Low As Reasonably Achievable ALI Annual Limit on Intake ANS American Nuclear Society ASTM American Society for Testing and Materials CM Controlled Access Area CAM Continuous Air Monitor CAR Corrective Action Report CEDE Committed Effective Dose Equivalent CFR Code of Federal Regulations Ci Curie cm Centimeter CoC Chain of Custody cpmCounts per Minute D&D Decontamination and Decommissioning DAC Derived Air Concentration dB Decibel DCGL Derived Concentration Guidelines DDE Deep Dose Equivalent DOC Decommissioning Operations Contractor DOT Department of Transportation DQO Data Quality Objective dpm Disintegrations per Minute DPW Declared Pregnant Woman 6

EPA U.S. Environmental Protection Agency FR Federal Register g

Gram GM Geiger-MOller HASP Health and Safety Plan HEPA High Efficiency Particulate Air HP Health Physics HSA Historical Site Assessment ISM Industrial Safety Manager I

Liter LCS Laboratory Control Standard LDE Lens Dose Equivalent LLRW Low Level Radioactive Waste m

Meter M&TE Measuring and Test Equipment MDA Minimum Detectable Activity MDC Minimum Detectable Concentration MDCR Minimum Detectable Count Rate mg Milligram mSv Milli-Sievert MSDS Material Safety Data Sheets MSL Mean Sea Level NCR Nonconformance Report NIOSH National Institute of Occupational Safety and Health NIST National Institute of Standards and Technology NUREG Nuclear Regulation NVLAP National Voluntary Laboratory Accreditation Program OSL Optically Stimulated Luminescence PTS Pneumatic Transfer Tube System QA Quality Assurance QC Quality Control 7

QOT Qualit Oversight Team OSHA Occupational Safety and Health Agency RAD Radiation Absorbed Dose RASS Remedial Action Support Survey RCRA Resource Conservation and Recovery Act Rem Radiation Equivalent Man RHWP Radiological and Hazardous Work Permits RP Radiation Protection RSC Reactor Safeguards Committee RWT Radiation Worker Training Rx Reactor SAR Safety Analysis Report SER Safety Evaluation Report SRPD Self-reading Pocket Dosimeter SS Stainless Steel SW-846 Test Methods for Evaluating Solid Wastes, EPA SW 846 TEDE Total Effective Dose Equivalent TLD Thermoluminescent Dosimeter TLV Threshold Limit Value TOCs Tops of Monitoring Well Casings.

TODE Total Organ Dose Equivalent TRIGA Training, Research, Isotopes, General Atomics TS Technical Specifications USAEA United States Atomic Energy Act USAEC United States Atomic Energy Commission USDOE United States Department of Energy USGS United States Geological Survey USNRC United States Nuclear Regulatory Commission VAMC Veterans Affairs Nebraska - Western Iowa Health Care System 8

I Executive Summary 1.1 Licensee/Site Name and Address Department of Veterans Affairs Nebraska-Western Iowa Health Care System Alan J. Blotcky Reactor Facility 4101 Woolworth Avenue Omaha, NE 68105 1.2 Site Description The Alan J. Blotcky Reactor Facility (AJBRF) is located in the Omaha Veterans Affairs Medical Center (VAMC) in the City of Omaha, Nebraska. The Omaha VAMC is built in a commercial area within the city limits and the reactor facility is housed in the southwest comer of the research building's basement. The reactor facility contains a low-power (20kW) Mark I Training, Research, Isotopes, General Atomics (TRIGA) nuclear reactor, which was operated as a source for neutron activation analysis of biological samples and for hot atom chemistry research. Additionally, beginning in 1989, the reactor was used for training Fort Calhoun Station power reactor operators.

1.3 Summary of Licensed Activities The AJBRF initial operating license (R-57) was issued on June 26, 1959 and the most recent license renewal was issued as Amendment Eleven on August 5, 2002. Amendment Nine was issued on April 12, 1991, to allow for operation of the reactor at steady state power levels up to a maximum of twenty kilowatts (thermal) and for the installation of a microprocessor-based neutron monitoring system.

1.4 Nature of Sources The radioactive materials licensed for use at the AJBRF primarily consisted of the reactor fuels and fission chambers which have since been transferred and sent for use in the United States Geologic Survey (USGS) reactor in Denver, Colorado. All licensed radioactive sources were removed and properly disposed, with the exception of Americium-Beryllium and Polonium-Beryllium sources.

9

See Table 4.1 for a list of radionuclides and maximum activities and quantities authorized under the current license.

1.5 Location and Storage of Licensed Radionuclides The AJBRF is located in the basement of an eleven-story medical center building erected in 1951. The AJBRF exterior walls are constructed of brick and reinforced concrete construction, including floors and ceiling, except that the walls between the rooms in the reactor area are of wood-stud plaster construction. All radioactive sources were maintained in a controlled storage area within the reactor facility or in the reactor pool itself.

1.6 Decommissioning Objective The objective of the AJBRF decommissioning is the release of the site for unrestricted use.

1.7 Derived Concentration Guideline Levels (DCGLs)

Site DCGLs and corresponding doses are based upon the results of the site characterization report and the guidance from the Nuclear Regulatory Commission (NRC) Decommissioning and Decontamination (D&D) computer code (as detailed in Section 6 of this report).

1.8 ALARA Evaluations Remediation activities within the AJBRF will be conducted to reduce the residual radioactivity distinguishable from background in accordance to 10 CFR 20.1402, Subpart E, so that the site can be released for unrestricted use. The guidance provided in Appendix D of NUREG 1727 was utilized to determine extent of the analyses required for license termination.

Removable surface contamination will be eliminated, where possible, by manual methods. A loose contamination limit of 1000 dpm/100 cm2 and a fixed contamination limit of 5000,dpm/100 cm2 shall govern the free release of materials. The 10 CFR 20 release exposure limit of 25 mrem/yr corresponds, at the AJBRF, to 27,000 dpm/1 00 cm2, as determined by the site characterization report (Attachment A). The calculated DCGL generally exceeds the contamination levels currently present on the structures and surfaces of 10

equipment. Plant systems and reactor components constitute an exception, and will receive decommissioning to meet the calculated DCGL.

1.9 Decommissioning Timeline The proposed initiation of decommissioning activities is early June 2005 with completion by February 2007, with final survey and site release, dependent upon NRC approval of this plan.

1.10 License Amendment for Decommissioning The Department of Veterans Affairs Nebraska-Western Iowa Health Care System, Omaha Division, AJBRF (License No. R-57) requests that its license be amended to incorporate the decommissioning plan and to modify the current license requirements for selected reactor surveillances.

11

2 Facility Operating History 2.1 License Status The AJBRF initial operating license (R-57) was issued on June 26, 1959 and the license was renewed as Amendment Eleven, issued August 5, 2002, authorizing operation for twenty years from that date. No subsequent amendments have been issued.

The reactor is a TRIGA MARK I reactor, owned by the U.S. Department of Veterans Affairs. Operated by the Omaha VAMC, the reactor is licensed pursuant to 10 CFR 50 (Domestic Licensing of Production and Utilization Facilities). The reactor was originally licensed for power at steady state power levels up to 18 kilowatts (thermal). Amendment Nine was issued on April 12, 1991, to allow for operation of the reactor at steady state power levels up to a maximum of 20 kilowatts (thermal) and for the installation of a microprocessor-based neutron monitoring system.

The reactor is a pool-type facility that was previously fueled with standard TRIGA fuel elements enriched to less than 20% uranium-235 zirconium hydride. Fuel elements were removed in June, 2002 and shipped to the USGS TRIGA reactor in Denver, Colorado.

2.2 License History Reactor construction began on January 8,1959. Construction of the AJBRF was completed on June 24, 1959 in conformity with the construction permit CPRR-36; the provisions of the Atomic Energy Act of 1954; and the regulations of the Atomic Energy Commission.

The AJBRF TRIGA reactor's maximum power from initial start-up in 1959 to October 2, 1995 was 18 kilowatts (kW). The reactor maximum power from October 2, 1995 to final shutdown on November 5, 2001 was 20 kW. The integrated power generated during operation of the AJBRF was approximately 515,058 kW-hrs.

12

2.2.1 Authorized Activities The AJB reactor was used to support nuclear medicine and research programs conducted at the medical center. Between 1959 and 1965 the reactor was funded as a national laboratory and employed approximately 30 people. The principal use of the reactor was for neutron activation of biological samples.

Typical irradiation times were up to 60 minutes in duration. Samples were irradiated in five ml vials. The vials were opened in the fume hood to allow the argon-41 (Ar-41) to vent to the atmosphere. The reactor was operated as a neutron source for neutron activation analysis of biological samples and for hot atom chemistry research. Additionally, beginning in 1989, the reactor was used for training Fort Calhoun Station power reactor operators.

2.2.2 Location and Storage of Licensed Radionuclides The AJBRF is located in the basement of the eleven-story Nebraska-Western Iowa Health Care System, Omaha Division. The AJBRF is constructed of brick and reinforced concrete construction, including floors and ceiling, except the walls between the rooms in the reactor area are of wood-stud plaster construction. The normal entrance to the reactor laboratory is through the door marked SW 2 (Figure2.1, Reactor Facility Floor Plan).

The TRIGA reactor is located in a below ground shield and pool structure with only vertical access to the core. The fission chambers and fuel resided in the AJBRF reactor pool until their removal and transfer to the USGS GSTR (Colorado) reactor in calendar year 2002. The below ground AJBRF TRIGA reactor contains a graphite-reflected fixed core resting on the bottom of a 6.1 m steel, concrete, and epoxy tank. The surrounding earth and demineralized water within the tank provided the required shielding so that no special containment building was necessary.

The materials licensed for use at the AJBRF primarily consisted of the reactor fuel and fission chambers that have been removed and sent for use by the USGS reactor in Denver, Colorado. Various other licensed radioactive sources were used for instrument calibration and reactor start-up and monitoring. All materials were present in solid form. As per, Figure 2.1, Reactor Facility Floor Plan, prior to removal, the fission chambers and fuel resided in the AJBRF reactor pool. All other sources on-site are stored in the general storage area, room SW2F. The following is a list of radionuclides and maximum activities and quantities authorized under the current license:

13

Table 2.1, Authorized Radionuclides Reactor fuels and fision chambers (removed to USGS): -

(56) Aluminum-clad TRIGA elements, eight percent by weight Uranium, enriched to less than 20% U-235 I (1) Stainless steel-clad TRIGA element, eight percent by weight Uranium, enriched to less than 20% U-235 1 (2) Fission chambers with.99 grams of U-235 1 Oth&ermaterials(remnaining'onr-site): '

Americium-Beryllium sealed sources (up to four curies)

Polonium-Beryllium sealed source (up to eight curies) 1 Sent to the USGS Geological Survey TRIGA Reactor (GSTR) facility.

All other sources were stored in the general storage area, room SW 2F (Figure 2.1, Reactor Facility Floor Plan) prior to their removal and subsequent disposal in January 2003. Samples to be irradiated were typically prepared in either room SW 2C or SW 2E. Isotopes were stored in the isotope-storage area SW 2F. The pneumatic transfer system is located at position indicated. Gamma counting was done in the area marked uShield' as shown.

14

Figure 2.1, Reactor Facility Floor Plan Medical Center Basement

\\ Cooling Pit Temporary HEPA Exhaust Ventilation Arrows Indicate evacuation route X

Dashed Ine Indicates operation boundary To first floorresearch 15

2.3 Previous Decommissioning Activities There has been no previous decommissioning activity performed at the AJBRF.

2.4 Spills and Burials Review of available documentation, including AJBRF operating records and annual reports, reveals no evidence of spills or other uncontrolled releases of radioactive materials, nor evidence of burials of radioactive materials. The results of the site characterization survey support this conclusion.

16

3 Facility Description 3.1 Site Location The AJBRF is located in the Omaha VAMC in the City of Omaha, Douglas County, Nebraska. Figures 3.1a-b are topographic maps of Omaha depicting the area surrounding the Omaha VAMC site. The reactor is housed in the basement of the southwest wing of the hospital (Figures 3.2 and 3.3) directly below the optometry clinic. The entire Omaha VAMC grounds are approximately 34.5 acres and the reactor facility occupies approximately 232.26 square meter (M2) of the hospital basement.

The Omaha VAMC site is situated approximately 4.83 kilometers (km) west of the Missouri River. The river level is normally at 267.9 meters (m) above Mean Sea Level (MSL) and the rolling hills in and around Omaha rise to as high as 361.2 m above MSL. Gently rolling topography surrounds the Omaha VAMC site with the medical center located atop a small knoll. The ground surface at the medical center varies in elevation between 365.8 m and 374.9 m above MSL.

These elevations represent some of the highest ground within the Omaha city limits, sitting approximately 83.8 m above the level of the Missouri River. There are no man-made or natural bodies of water on the hospital grounds.

To the west of the Omaha VAMC, across 42nd St., is a residential area of single-family detached homes. To the south, across Center Street, is a commercial district of small businesses and restaurants. To the east is a golf course (the Field Club of Omaha), and to the north, across Woolworth Avenue is the Douglas County Hospital.

There are no industrial activities in the area that will be impacted by the facility's decommissioning. Approximately 3.2 km northwest of the site is a large railroad yard and 12.9 km to the northwest is Offutt Air Force Base, Headquarters for the Air Combat Command of the U.S. Air Force. The Omaha airport is more than 9.7 km from the facility, and a low altitude airway (914.4 m to 5,181.6 m MSL) passes near the site. The nearest interstate highways (1-80 to the south and I-480 to the east) are more than 1.6 km from the facility.

The AJBRF site is part of a functioning medical center, including support buildings, as described in Figure 3.2, which shows the physical features of the hospital site. Figure 3.2 also provides a plan view of the hospital facility and relative location of the reactor facility.

17

The normal entrance to the reactor laboratory is through the door to room SW 2 (Figure 2.1). The area to the left of the access door serves as a health physics control point. Samples to be irradiated were typically prepared in either room SW 2C or SW 2E. Isotopes were stored in the isotope-storage area SW 2F. The pneumatic transfer system was used to send samples through the core. Gamma counting of irradiated samples was done in the area marked shield.

The functions of the areas/rooms in the AJBRF are as follows:

Table 3.1, Functions of Areas/Rooms Room or Area Number :-UX.

Use SW 2 Radioisotope Reactor Research Lab SW 2A Office and Electron Microscope SW 2B Office, Dark Room, Clothing Change Room, and Intermediate Monitoring Room SW 2C Nuclear Research Lab & Office SW 2D Walk-in Cooler SW 2E Nuclear Research Lab & Office SW 2F Isotope Storage & General Storage Note: The following rooms are considered to be outside the AJBRF boundary and are included for information only.

Table 3.2, Included rooms outside of the AJBRF SW I Lockers SW 1A Restroom and Shower SW 3 Storage Room and Temporary Frisking Area The reactor room ventilation supply provides 100% outside air (heated or cooled) to the reactor laboratory at the rate of 43 cubic meters per minute (m3/min) through six ceiling ducts. The exhaust effluent of 84.1 m3/min exits the reactor room into the outside air by means of an exhaust fan installed in the outside wall of the building. In addition, two laboratory fume hoods (Figure 3.3) were operated continuously and exhausted a total of 26 m3/min by means of fans installed on the roof of the medical center (Figure 3.3). Since the blower for the hood exhaust is on the roof, the entire duct has a negative pressure relative to the adjoining hospital area. Any leakage would flow into the duct, thus eliminating the potential for exposure within the medical center. The ventilation exhaust point of interest is the area labeled Reactor Lab Hoods on Figure 3.3. As discussed further in Section 11 of this plan, these hoods will be secured and 18

made non-operational during the D&D activities We don't have to be there. The plan is for you to wake up Tuesday morning, and to notice that we don't have a couch. You will then go to work. When you return, there will be a couch where once there was not.

The reactor is located near the bottom of a cylindrical pit 6.1 m below ground level. The pit contains a steel tank with an inner diameter of 208 cm and a wall thickness of 64 cm. The tank rests on a 0.28 m concrete slab with approximately 0.25 m of poured concrete surround the outside of the tank. The inside of the steel tank is covered on the sides by a layer of gunite (approximately 0.05 m thick) and on the bottom by poured concrete (approximately 0.1 m thick). The entire inner surface is coated with two applications of a waterproof epoxy resin.

The reactor pit was designed to ensure against water leakage. The gunite and its waterproof coating protected the steel tank against corrosion by water. If a small defect in the coating should occur, the steel tank provided a secondary containment vessel. Approximately 4.9 m of water served as shielding above the reactor core. Figure 3.4 provides a schematic of the reactor and pit.

Three emergency storage pits were located immediately adjacent to the reactor tank. The pits were vertical steel pipes 0.25 m in diameter and 3.05 m long. The pits were intended to store irradiated specimens or failed fuel elements.

The core structure is bolted to the bottom of the reactor pit and is approximately 0.36 m long. The core contained 85 fuel-element positions; 57 contained active fuel elements, and the remainder of the positions were occupied by graphite elements (elements in which the uranium-zirconium-hydride fuel is replaced by graphite). The assemblies were supported on the top and bottom by grid plates.

Both grid plates were 0.019 m of aluminum. The final core consisted of the original 56 aluminum-clad fuel elements and one stainless steel clad element that was added to the core on October 2, 1995. The core was cooled by natural circulation of water, flowing through the core from bottom to top.

A cylindrical reflector that was 0.31 m thick (inner diameter of 0.43 m and outer diameter of 1.07 m) surrounded the core. The graphite reflector was completely encased in a welded aluminum can. Reflectors on the top and bottom of the core were 0.1 m graphite sections encased in fuel element cans.

Irradiation facilities were part of the reactor structure and provided for the production of radioisotopes. These include a rotary specimen rack located in the well in the reflector can, a pneumatic transfer tube, and a central thimble. In addition, odd-shaped specimens could be irradiated in the water outside the reflector.

The rotary specimen rack (also referred to as the "Lazy Susan") is an aluminum ring that rotated around the core. Forty aluminum cups, evenly spaced, were hung from the ring and served as irradiation specimen holders. The Safety Analysis Report (SAR) contains additional detail about the rotary specimen rack.

19

The pneumatic transfer system provided for the production of isotopes with short half-lives. It consisted of two tubes leading down through the water tank to a position at the outer edge of the core, where the tubes were joined. A blower connected to the other tube provided the pressure difference to inject or eject the specimen. Specimens were inserted and removed from the pneumatic transfer system in the reactor laboratory.

The central thimble was installed to provide irradiations or experiments in the region of maximum neutron flux. The thimble consisted of a vertical aluminum tube (with an inner diameter of 0.034 m) leading from the top of the reactor pit through the center of the core and ending at the bottom of the core. The vertical aluminum tube had holes in the region of the core to allow shield water to be removed from the portion of the central thimble above the upper grid plate using air pressure. This provided a highly collimated beam of neutron and gamma radiation for experiments. The central thimble was only used once during reactor operation, for the determination of dose levels.

Three boron carbide control-rods were operated in perforated aluminum guide tubes. Each control rod had an extension tube that connects to the control-rod drive mechanism. The control-rod drive mechanism is located on the bridge at the top of the reactor pool.

The reactor was cooled by natural convection of the pool water. A five-ton freon vapor-compression chiller with an air-cooled condenser was used as the heat sink. Water from the reactor went through a water monitor where the temperature, gamma activity, and conductivity of the water were measured. The water was first pumped to the chiller unit, then through a filter and a mixed-bed demineralizer before returning to the tank.

A wide-range fission chamber and a boron-lined uncompensated. ion chamber provided the reactor core monitoring system. The fission chamber was removed in December 2002 and transferred to the USGS GSTR facility in Denver, CO.

The following facility and system modifications were performed, based on the ABJRF annual reports and Reactor Safeguards Committee (RSC) minutes and logs back to 1959.

• In 1966, the rotary specimen rack was replaced. The original rotary specimen rack is stored along with another irradiated component in a below ground location accessed from the outside noted on Figure 2.1 as SW 2D.

• In the 1960s, the pneumatic transfer system, which was originally routed to the first floor, was cut and re-routed to the reactor laboratory.

• In 1995, a stainless steel TRIGA fuel element was added to the core on October 2, 1995 due to the core fuel bum-up since 1959.

20

• In 1995, the terminus end of the pneumatic transfer system tube was replaced due to leakage.

• In 1999, the original 1959 General Atomics console was replaced with a solid state General Atomics Mark II console along with safety, shim, and regulating rod drives.

21

I Figure 3.1a, USGS topographic map of Omaha, NE 22 co\\

I-Figure 3.1 b, USGS topographic map of Omaha, NE (detail)

I I iDA I

I I011 F1X'

%\\

I i

I 1

w 3 L1k77JFT'-

I 9 I 

I 

mra

" qZ "Sm F-4 j K, I 1 " m

. T.1 liqm k~

-~ ;

T'Tri 17C 11 "sVC77 LiI

/ i IdI t

M 4-MAt I-JU/

11 11 s

~~I W

=f1, e-i riz TA r 7

-76

^

A{ t EMW VY Jr II w

i I

4 JE Vu' i~zR#

462 i

I 17 2

AU t

I'U"i t'

wrr=

=_

).

j wt't I"Y I

e

(£ Z-II 4

I L,

.M

-F A A x

'N 1/V'yI&I ;

s rl N1-1 if r



iii r.'i f FM E

WAAz i

-. 1WS4&

I'

-. I

!~I!

or3 X Xff 0

CM M A A11 L

f 1W 0

IrAW'vxtrrff NsNV irf I

f,;C77 VA

.A-rM 6t l M;97I x-4 I

I I f-

,-l F

}1.>

.s--

N Alt v

I

I J/ - IT' p

P, I

I t

5L F

Ita W7

,IF

.s

• 1 IE kkilmt Si-I MLIMWISI
  

V~ J~-A i@

I S ; I

41

I~ I

_Kw U-I-

_u" E

I j71 L ^1 L W, I K,7
I tX I

0

79 I

PM __jYfW'_

I

w I.A 74EM I s r

i~ B~J I H 

4 

I I

1

I0P0i k

t51 tW 'e 3&!

KU4 P.-"

.K I

/

.WA..FAWhdLLMr-r*~l X

fl9 M-.

I I

t

• 1\\\\ I,J El III VI Ez

>,'I JL M N U 1, 1 U At-L:,

WL&b M3V jiX AM!

1\\M.

X j -e f QI f ! I 4~

WLL-1 EA-1 r v 0 o 23 CGUL

Figure 3.2, Site Plan of the Omaha VAMC T) caz QN Center Street c'O3 24

Figure 3.3, Omaha VAMC Roof Exhaust Locations Nuclear Medcine Sulte 1 Exhaust Elev. 371'- O 25

Figure 3.4, Reactor and Pit 26

3.2 Population Distribution Female persons, percent, 2000 51.1%

Persons of Hispanic or Latino origin, percent, 2000 6.7%

White persons, not of Hispanic/Latino origin, percent, 2000 78.2%

High school graduates, persons 25 years and over, 1990 221,316 College graduates, persons 25 years and over, 1990 65,087 The Omaha metropolitan area includes suburbs in both Nebraska and western Iowa and has a population of 716,998 as of 2000. The city of Omaha had a 2000 population of 390,007 and covers 307.9 square km (kM2) in area. Table 3.3 displays the 2000 census data for Douglas County, Nebraska as distributed by population segments.

Table 3.3, Douglas County, Nebraska 2000 Census Data Population, 2000 463,585 Population, percent change, 1990 to 2000 11.3%

Persons under five years old, percent, 2000 7.4%

Persons under 18 years old, percent, 2000 26.6%

Persons 65 years old and over, percent, 2000 11.0%

White persons, percent, 2000 81.0%

Black or African American persons, percent, 2000 11.5%

American Indian and Alaska Native persons, percent, 2000 0.6%

Asian persons, percent, 2000 1.7%

Native Hawaiian and Other Pacific Islander, percent, 2000 0.1%

Persons reporting some other race, percent, 2000 3.4%

Persons reporting two or more races, percent, 2000 1.8%

The Omaha VAMC has a staff of approximately 888 people and has a typical daily load of 250 outpatients. There are 100 inpatient beds with an average daily inpatient census of 72.

27

In light of the proposed decommissioning goal and associated schedule, no significant population shifts are anticipated during decommissioning activities.

This assertion is based on local knowledge relevant to development, growth, and pattern of population growth, which increased from 1990 to 2000 by 11.3%.

3.3 Current and Future Land Use The AJBRF is located in the Omaha VAMC, which is presently, and will continue to function as, a medical center facility operated by the Department of Veterans Affairs. The areas surrounding the reactor in the basement are used as laboratories for medical research. It is anticipated that the basement area currently housing the AJBRF will be turned into office space for hospital staff or for use as laboratory space.

3.4 Meteorology And Climatology The Omaha climate is typical of the Midwest region of the U.S. with relatively warm summers and cold, dry winters. It is situated midway between two distinctive climatic zones - the humid east and the dry west. Fluctuations between these two zones produce periods of weather conditions that are characteristic of either zone or combinations of both. Omaha is also affected by most storm systems that cross the country. The prevailing winds at the Omaha Airport (Eppley Airfield) are south-southeast during most of the year, shifting to north-northwest during the winter season, with a mean wind speed of approximately 16.09 km per hour (kph).

Most of Omaha's precipitation falls during sudden showers or thunderstorms from April to September. Of the total precipitation, about 75% falls during this six month period, predominantly as evening/night showers or thunderstorms.

Although winters are relatively cold, precipitation is light, with only ten percent of the total annual precipitation falling during the winter.

The mean date of the last freeze in spring is April 14th, and the mean date of the first freeze in autumn is October 20th. The longest freeze-free period on record is 219 days in 1924, and the shortest period is 152 days in 1885. The average length of the freeze-free period is 188 days.

Omaha and the surrounding region are considered an Attainment Area relative to all airborne pollutants and thus meet the National Ambient Air Quality Standards for those pollutants.

Table 3.4 and Figure 3.5 describe the tornado locations for the state of Nebraska and the area surrounding the Omaha VAMC. Tornadoes have been recorded in the general area of the site and have caused minor damage to the hospital; however, no damage has occurred to the AJBRF. The AJBRF is in the 28

basement of the Omaha VAMC surrounded by poured concrete walls with no windows and with ten cm of concrete overhead. Because of this reinforcement, the area in and around the reactor room has been designated as the Omaha VAMC's tornado shelter.

29

Table 3.4, Tornado Frequency Data Year Tornadoes Deaths; Injuries 1970 14 0

1 1971 52 0

1 1972 30 0

0 1973 19 0

13 1974 32 0

20 1975 78 4

141 1976 26 0

23 1977 68 0

1 1978 42 0

9 1979 20 0

1 1980 38 5

204 1981 19 0

0 1982 34 0

0 1983 15 0

0 1984 50 0

24 1985 52 0

7 1986 54 0

9 1987 26 0

2 1988 20 2

1 1989 41 0

3 1990 88 0

10 1991 65 0

6 1992 75 0

5 1993 70 0

18 1994 55 0

3 1995 26 0

0 Source: The Disaster Center, Nebraska 30

Figure 3.5, Omaha Historic Tornado Touchdown Map 31

3.5 Geology The City of Omaha lies within the Dissected Till Plains of the Central Lowland Psysiographic Province of the United States. The Omaha VAMC is located on a knoll surrounded by gently rolling topography. The medical center sits at an elevation of approximately 366 m above Mean Sea Level (MSL) in a commercial area within the city limits of Omaha. The City of Omaha sits at an elevation of 304 m to 396 m above MSL and is bordered on the east by the Missouri River, separating Nebraska from Iowa. Thus, the hospital is on some of the highest ground within the city.

The surface soils in the Omaha 'area are primarily loess and glacial drift deposits.

Two stages of glaciations, the Nebraskan and the Kansan, left thick deposits of till overlying bedrock. It is believed that much of the glacial till has been eroded in the vicinity of the Omaha VAMC and that not more than 30 m remains. The till consists mainly of lean and gravelly clays with a few lenses of sand gravel. The exact depth to bedrock directly below the Omaha VAMC site is not known, but is estimated to vary between 305 m and 320 m MSL, on the basis of the nearest top bedrock information.

The loess at the site is of Peorian and Loveland Formations of the late Pleistocene Epoch. The soil classification of the Peorian indicates that the material consists predominantly of clayey silts and lean clay. The soil of the Loveland formation varies from clayey silt to fat clay with minor amounts of sand and clayey sand in the basal part of the formation. At the Omaha VAMC site, the Peorian is from 9.1 m to 13.7 m thick and the Loveland is over 18.3 m thick.

Therefore, the total thickness of the overburden is between 55 m and 65 m.

Soil boring performed during hospital construction showed the topsoil layers (depth of 7.6 m) to be Peorian Loess, a wind-blown deposit of clayey silt having low plasticity. The next meter consists of Loveland Loess, a windblown deposit of silty clay having medium plasticity. The bottom 0.30 m of the boring was in glacial clay having a higher plasticity. The test data shows the soils to be strong near the surface, then decreasing to medium strength at a depth just above the very strong clay layer. No water table was encountered.

Bedrock in this area is limestone and shale of the Pennsylvania period. The surface of the bedrock is very irregular because of erosion that followed the uplift of the area in early Pennsylvania time and continued on to the Pleistocene period. This uplift brought the granite to within 183 m of the surface in certain areas, forming a ridge known as the Nemaha Ridge or Arch. Also, extensive faulting occurred that developed a major fault, known as the Humboldt fault, which has a throw of over 274 m. There is no evidence of activity along this fault in recorded time, and it has not been considered to be a capable fault within the guidance contained in 10 CFR 100.

32

3.6 Surface Water Hydrology There is no surface water located on the grounds of the Omaha VAMC. The closest bodies of water to the reactor site are the Big Papillion Creek, approximately three km to the west, and the Missouri River, approximately five km to the east (Figures 3.1a-b).

3.7 Groundwater Hydrology As seen in the area map (Figures 3.1a-b), the Big Papillion Creek, which runs in a southeasterly direction, is approximately three km west of the site. With the water table troughing along this creek, the underground water would migrate along the creek until it returns to the Missouri River South of Offutt Air Force Base (13 km from the site). The Missouri River is approximately five km due east of the site; however, the topography is such that any transport of radioactive materials by groundwater would be towards the Big Papillion Creek.

The closest wells in the immediate area are located 2.5 km due west on the Aksarben grounds (Figures 3.1a-b), and in an area near 84th and L Streets.

Regarding the Aksarben site, the well would not be subject to contamination since the water flow occurs to the southwest. The site at 84th and L Streets is more distant than the Big Papillion Creek, which would be the major source of migration of radioactive materials to the environment.

On the basis of logs of the borings drilled in 1946, the zone of saturation was believed to be below 19.81 m. Test wells drilled during the site characterization in November 2002 (see Figure 3.6 for drilling locations) determined water levels to be from 20 - 22 m below grade. Based on the three borings performed the direction of groundwater flow is generally west by southwest with a hydraulic.

gradient of 0.01 meter/meter. Since the medical center is sited on a knoll, there is reasonable assurance that decommissioning activities at the facility will impact neither surface nor groundwater. However, the test wells will be maintained and monitored throughout the decommissioning and final survey. The well details and discussions of the boring results are provided in the site characterization report (Attachment A).

33

Figure 3.6, Test wells, relative elevations of the groundwater, and resulting hydraulic gradient.

EB -2 (29.90')°

& EB-1 (30.48')

RESEARCH BLD I = 0.01 Scle: 1/32'" -

'0' 0N

&EB-3 (31.33')

'-9.

r

{

I

§ Direction of Croundwoter Flow (90-I Monitoring Well

_-w-Water Table Contour Une (feet:

(29.9s) Relative Elevation of Water Table (feet) i - 001 Hydroulic Grodient (ft/It)

Groundwater F A.J. Blotcky Ri Veterans Affair HDR Engineering. Inc. Omaha. Nebraska

'low Direction (Nov. 25,2002) aactor Facility i Medical Center Dote Dec 2002 1

34

3.8 Natural Resources There are no commercially or recreationally significant natural resources in the vicinity of the site. A telephone conversation on July 18, 2002, with Mr. Emmitt Egr of the Papio-Missouri Natural Resources District, verified that there are no natural resource, ecology, or endangered species considerations at or near the Omaha VAMC site.

35

4 Current Radiological Status of the Facility Duratek, Inc. performed a site characterization report of the AJBRF in December 2002 to ascertain the nature and extent of radiological activities and to facilitate the preparation of the AJB Decommissioning Plan and Cost Estimate (Table 15.1). The results indicate contamination only inside the facility. The site characterization report serves as Attachment A of this document. The site characterization survey of the areas outside the containment building found no detectable radioactive contamination above background levels. The highest concentration of radiation exposure was found in the south lab hood in room SW2C and in drain three in the south lab hood of room SW2D.

Table 4.1 provides a list of radionuclides from 10 CFR 61, Tables I and 2, which are considered to be potential nuclides of concern at the AJBRF. Additional sampling was performed to further identify and qualify radionuclides of concern at the AJBRF.

Table 4.1, Potential Radionuclides of Concern at the AJBRF

' Nuclide Half Life Major Radiations a,

Energies (MeV) and intensities (/)'

Alpha Beta (average)

Gamma H-3 12.28 yrs 5.685 keV 100%

C-14 5730 yrs 49.47 keV 100%

Mn-54 312.7 834.8 keV days_99.98%

Fe-55 2.7 yrs Low energy x-rays Co-57 270.9 122.06 keV days 85.51%

Ni-59 7.5 E4 yrs Low energy x-rays Co-60 5.27 yrs 95.79 keV 1173 keV 100%

100%

1332 keV 100%

36

Table 4.1, Potential Radionuclides of Concern at the AJBRF (con't)

Nuclide'

-Half Life Major Radiations

,,l

,Energies (MeV) and intensities (%)j '____

Ni-63 100.1 yrs 17.13 keV 100%

Zn-65 244.4 1116 keV 50.75 days Sr/Y-90 28.6 yrs 195.8 keV 100%

934.8 keV 100%

Nb-94 2.03 E4 145.8 keV 702.6 keV 100%

yrs 100%

871.1 keV 100%

Tc-99 2.13 E5 84.6 keV 100%

yrs Ag-II Om 249.8 66.6 keV 657.7 keV days 98.4%

94.4%

884.7 keV 72.6%

1-129 1.57 E7 40.9 keV 100% 39.58 keV yrs 7.5%

Cs-I 34 2.06 yrs 156.8 keV 100% 604.7 keV 97.6%

795.8 keV 85.4%

Cs-137 30.17 yrs 156.8 keV 661.2 keV 94.6%

85.1%

415.2 keV 5.4%

Eu-I 52 13.6 yrs 300.8 keV 121.8 keV 28.4 27.8%

964 keV 14.4%

1085.8 keV 10%

1112 keV 13.3%

1407 keV 20.7%

344 keV 26.5 %

778.9 keV 12.7%

37

Table 4.1, Potential Radionuclides of Concern at the AJBRF (con't)

Nuclide, HalfLife:

Major Radiations, Energies (MeV) and intensities(%)

Eu-1 54 8.8 yrs 225.4 keV 100% 123.1 keV 40%

1274 keV 35.5 Eu-1 55 4.96 yrs 45.2 keV 100%

86.5 keV 31 %

105.3 keV 20.7%

Pb-210 22.26 yrs 6.5 keV 100%

Low energy photon Th-230 7.7 E4 yrs 4670 keV 100%

U-234 2.45 E5 4724 keV yrs 27.4%

4776 keV 72.4%

U-235 7 E8 yrs 4396 keV 55%

143.8 keV 10.5%

185.7 keV 54%

U-238*

4.51E9 5150 keV 103keV 21%

63 keV 3.5%

yrs 25%

193 keV 79%

93 keV 5.4%

4200 keV 75%

Pu-239 2.41 E4 5104 keV yrs 11.5%

5142 keV 15.1%

5155 keV 73.3%

Pu-241 14.4 yrs 5.23 keV 100%

Am-241 432 yrs 5443 keV 13%

59.5 keV 35.9%

5486 keV 85%

• The beta particles and gamma photons for U-238 are from the Th-234 daughter.

38

Table 4.2 provides a list of radionuclides of concern based on samples taken from the demineralizer resin.

Table 4.2, Resin Sample Analysis Results Summary

., Radionuclide,

-Result

,Uncertainty,'.

MDA PpCIlg

-pCg pCVg' Ag-I IOM 2.39E-01 2.16E-01 3.51E-01 Am-241 1.36E-02 4.1 OE-02 1.03E-01 C-14 4.OOE+02 5.94E+00 3.22E+OO Cm-244 3.70E-02 4.89E-02 7.52E-02 Co-57 1.02E+00 1.81E-01 I.IOE-01 Co-60 2.61E+01 2.03E+00 3.36E-01 Cs-134 8.51 E-02 1.75E-01 2.96E-01 Cs-I 37 5.37E-01 3.52E-01 3.74E-01 Eu-152 5.33E+00 9.91E-01 1.08E+OO Eu-I 54 2.29E-01 4.07E-01 7.49E-01 Eu-I 55 2.53E-01 2.08E-01 3.48E-01 Fe-55 2.95E+01 9.73E+02 1.36E+02 Gross Alpha 9.60E-01 4.89E-01 6.14E-01 Gross Beta 2.42E+01 1.47E+00 8.48E-01 H-3 7.79E+00 2.29E+00 3.58E+OO 1-129

-5.72E-02 1.97E-01 3.26E-01 Mn-54 1.54E-01 2.42E-01 4.23E-01 Nb-94 3.91 E-02 2.65E-01 4.56E-01 Ni-59

-6.00 E+00 1.39E+01 1.80E+01 Np-237 2.33E-02 5.69E-02 1.30E-01 Pb-210 3.79E+00 7.97E-01 9.95E-01 Pu-238

-8.03E-03 9.37E-03 9.49E-02 Pu-239 2.88E-02 4.51 E-02 7.28E-02 Pu-241 5.61E+OO 2.96E+00 5.IIE+OO Pu242 O.OOE+OO O.OOE+O0 5.85E-02 39

Table 4.2, Resin Sample Analysis Results Summary (con't)

Radionuclide Result Hi Uncertainty -

_v -;,MDA--

.;'p Cilfg

- ; ' Ci/g,.

p Cilg Pu-244 2.16E-02 4.34E-02 5.86E-02 Tc-99 1.01 E+00 9.81 E-01 1.64E+00 Th-228

-2.65E-04 2.79E-02 9.44E-02 Th-230 2.64E-01 1.25E-01 6.92E-02 Th-232 1.91 E-03 2.73E-02 8.92E-02 Total Sr 3.29E-01 4.12E-01 6.92E-01 U-234 1.05E-02 2.58E-02 5.87E-02 U-235

-2.66E-03 5.34E-03 7.24E-02 U-236 O.OOE+00 O.OOE+00 3.81 E-02 U-238 2.31 E-02 3.62E-02 5.85E-02 Zn-65 1.61 E+00 7.62E-01 9.39E-01 Bolded results are those that exceeded the analysis MDA.

While estimates of the radioactivity inventory can be determined 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 incident neutrons, direct measurements are generally more reliable. Direct measurements will be used during actual removal and/or dismantlement of components. These measurements will provide further detail on the basis for specifying the necessary safety measures and procedures for the various dismantling, removal, contamination, waste packaging, and storage operations so that exposure to personnel is maintained as low as reasonably attainable (ALARA).

4.1 Contaminated Structures Structures at the facility where licensed activities occurred that contain quantities of radioactive material in excess of site background levels include:

• Reactor Room Console Area;

• Room SW2C;

• Room SW2D; and

• Room SW2F.

40

Please refer to Figure 2.1 for a drawing showing the locations of radionuclide material contamination.

The following Tables provides a summary of the background and contamination levels identified during the site characterization report:

Table 4.3, Summary of Background Levels Survey Description Measurement Type Maximum/Average Designation/

radionuclide activity Location present (dpm/1OOcm2) in;,

each room, B0011/Ceiling

Background

Total Beta 787/560 Tiles Basement contamination B0009/Painted Background Total Beta 934/635 brick Rm. B003 contamination B0009/Painted Background Total Beta 908/577 Brick in Rm.

contamination B003 B0010/Bare

Background

Total Beta 1,404/957 Brick in Rm.

contamination B003 B0001

Background

Total Beta 714/409 Concrete contamination B0003/Sealed

Background

Total Beta 1,286/736 Concrete contamination Sidewalk 22-30/Concrete Resurvey of Total Beta 964/558 Sidewalk Pneumatic contamination transfer system.

Background

B9999ITile

Background

Total Beta 387/142 Floor Basement contamination B0006//Tile

Background

Total Beta 1,454/942 Walls contamination 41

The Table below summarizes the locations and extent of radioactivity detected within the AJBRF relative to structures, systems, and equipment.

Table 4.4, Locations and Amounts of Contamination Levels Room.-.'

Numbers.'

located in the facility Description of. Room '. :> '

Activities. ' :.

...,i..

W.

.r

. 4.

-I A..

. Jo,

.S.

^.

i,-,'-

,'.1

.s '

t

. fl

':, ',',. fl ' ' -

1,

),. ' 'I s,'

A..

.. './e.

i.'"

i; t,'X' X, a 0'. '.

-'t '

a,\\-',

F # Is ". ' "

.s

'Locations of

.tion in each.

'room (i e.,

walls, floors, wall/floor joints,'. '

structural '.':

steel surfaces, -..

ceilings;.etc.):

Maximum/

Average radiation.-...

levels present

.(mrem/hr) or

.(urem/hr) at each location Maximum/'

Average radionuclide activity present...

(dpm/1 00cm2)

-in each room Mode of.'",'.

contamination for'..

,each surface (Direct Scans)

SW 2 Reactor Fuel Pool 10-12 pr/hr 12,116/313 Surface Room Cover contamination Console due to research Area operations SW 2 Reactor East Cooling 10-12 pr/hr 1,734/508 Due to sample Room Pit Floor handling and Console research Area operations SW2C Nuclear Walls 1-4 22 pr/hr/

2,227/241 Surface Research 6-10 pr/hr contamination Lab and due to research Office operations SW2F Source Concrete 8-16 pr/hr 15,682/2,071 Due to source Storage Floor handling and Room authorized drain discharge SW2F Source Wall 600 pr/hr 8,748/226 Source handling Room Penetrations 1-64 42

4.2 Contaminated Systems and Equipment The following table describes the types and activity of radioactive material detected in the facility systems and/or equipment, except for internal components of the reactor.

Table 4.5, Types and Activity of Radioactive Material Contamination

....... s s...

Systems or ':

Equiprnent ;-

A;

^

A...

.^

.. v He..

i-Maximum!..

Average.

radionuclide activities:;: '

(dpm/1OOcm 2)

Maximum/Average,;,.'.

radiation levels at the surface of each piece of equipment (mrem/hr) or (urem/hr)

Mode'of Contamination for each Sufface6--

(Direct Scans)

Pneumatic 1,243/304 General Area near Inside surface due to transfer system, system components passage of samples to Rm. SW 2

<1 mrem/hr and from the reactor (Reactor Room Console Area)

Room 72,211/14,481 General Area near hood Exhaust of SW2C/Lab Hood

<1 mrem/hr gases/particulates generated during sample preparations and analysis Lab Hood Drain 1,554,491/

General Area above Passed practices of D03/ Rm. SW 2D 1,554,491 drains <1 mrem/hr discharging samples as allowed by federal regulations and operating license Sink Drain/ Rm.

9,070/9,070*

General Area above Passed practices of SW 2D drains <1 mrem/hr discharging samples as allowed by federal regulations and operating license Sink Drain/ Rm.

2,017/2,017*

General Area above Passed practices of SW 2C drains <1 mrem/hr discharging samples as allowed by federal regulations and operating license

• Only one measurement was taken.

43

4.3 Reactor Pool and Components The following Figure depicts radiation levels associated with internal reactor components. These values represent exposure rates at the top of those components. Historically, the activity of the reactor coolant has been extremely low and recent reactor pool water samples indicate non-detectable activity.

Therefore, loose contamination levels should be minimal. Actual radiation levels will be determined during decommissioning planning and prior to removal of components allowing for proper implementation of controls. Figure 4.1 provides a gamma dose rate profile of the reactor components.

Figure 4.1, Reactor Dose Rates Control Port f zSusan

/ Lead Brick \\

0 1 w-Shim fety Rod 6

Cen r Thimble 0.7 Lazy Susan Hou Ing

.a 1.1 mS afety 0.5 egulating R d

1. Gamma Dose Rate in R/hr 44

4.4 Surface Soil Contamination An environmental analysis of surface soil samples was performed to determine the presence of radiological or hazardous contamination present outside of the containment building. None of the soil samples analyzed by gamma spectroscopy showed any measurable radioactivity. Presently, there is no surface soil that has been negatively affected by reactor operation.

A background study of soils was not performed because no radioactivity was detected in the characterization soil samples. If radionuclides attributable to licensed activities (e.g. Cs-137) had been identified in significant quantities in the characterization soil samples then a background study would have been performed on soils, sediments, and waters with common characteristics to the samples and measurements collected on site but that were unaffected by reactor operations.

4.5 Subsurface Soil Contamination An environmental analysis of subsurface soil samples was performed to determine the presence of radiological or hazardous contamination present inside and outside of the containment building. Subsurface samples were taken from locations adjacent to the reactor vessel inside the building and from the outside of the building below the vessel. None of the soil samples analyzed by gamma spectroscopy showed any measurable radioactivity. Presently, there is no subsurface soil that has been negatively affected by reactor operation.

Figure 4.2 indicates the location of soil borings performed in the soil adjacent to the reactor pit. Note that soil-boring number IB-4 was dug from outside the facility to allow for access under the reactor pit. In addition, Figure 3.6 shows the location of soil borings outside the facility.

45

Figure 4.2, Soil Boring Locations N

A Scale: 1/160 = IV" Levend Operston Doundosy O

Soll Uamptling Boring

_ Direction of AnSub DlaPring Note?

134 was drlled at an angel of ar frOm the vewtlcal to rocover Soil from diroetly beneath the rpactor.

Medlcal Center Basement

• 1 L

r.-- ;

toto is-2 &

I

&I o

13-11

!3-.3I r

I Il Research Building

- H Sol Sampling Borlngs A.J. Blotcky Reactor Facillty Veterans Affalr Medical Center HDR Englineftn.

1.

Omacn~o, NblobS'4c JFr. 2OG3 2

46

4.6 Surface Water Surface water does not exist at the Omaha VAMC, thus no surface waterl has been negatively affected by reactor operation.

4.7 Groundwater An environmental analysis of groundwater was performed to determine the presence of radiological or hazardous contamination present outside of the containment building. Presently, there is no groundwater that has been negatively affected by reactor operation.

47

5 Dose Modeling Evaluations As described in various sections of this plan, the reactor facility and affected outside areas of the facility are small in nature and do not contain wide-spread nor significant levels of residual radioactivity. Accordingly, the site is considered to have low residual concentrations of radioactivity present that are concentrated in small areas of the facility and not wide-spread though out the facility. As part of the decommissioning planning per ALARA, radioactivity levels will be reduced in all areas to values far below the unrestricted release criterion of less than 25 mrem/year. Upon achieving this dose objective, development of complex dose modeling is not warranted. Furthermore, as part of the site characterization effort, site-specific DCGL (27,000 dpm/100cm2) for meeting unrestricted release criteria were determined using appropriate computer codes and standards.

Conservatively, decommissioning is being planned to decontaminate and remove radioactive concentrations at the facility to values in the not to exceed a range of 5,000 dprnl100 cm2.

5.1 Unrestricted Release Using Screening Criteria As part of the site characterization process, DCGLs for unrestricted use of the facility were established by analyzing samples from the reactor demineralizer system that are considered to be representative of the types of radionuclides present. To support the establishment of the site-specific DCGLs NRC D&D screening codes were utilized, as were radionuclide-screening values contained in the Federal Register (FR Vol.63, No. 222, 11/18/98). Additional details regarding the DCGLs are documented in the attached site characterization report (Attachment A).

5.1.1 Building Surfaces Exterior building surfaces do not contain concentrations of residual radioactivity concentrations. Accordingly, the need to perform dose modeling other than what already has been completed in the process of establishing DCGLs for the site is not a consideration and appears not to be warranted.

48

5.1.2 Surface Soil Based on characterization results for the soils associated (inside and outside) dose modeling is not needed since the survey results reveal no presence of radioactivity above background values.

5.2 Unrestricted Release Using Site-specific Information The dose modeling methodology requirements necessary to satisfy this option is not considered applicable to this facility as discussed above.

5.3 Restricted Release Using Site-specific Information The dose modeling methodology requirements necessary to satisfy this option is considered not to be applicable to this facility as the objective is for unrestricted release of the facility for use.

5.4 Release Involving Alternative Criteria This dose modeling methodology for this criterion is not considered applicable, as the goal is to meet unrestricted use criteria for the facility.

49

6 Alternatives Considered and Rationale Supporting the Decommission Strategy Selected 6.1 Decommissioning Objective and Strategy Selected The objective of the AJBRF decommissioning is the regulatory release of the reactor site and adjacent areas to unrestricted use conditions according to 10 CFR 20.1402. The facility which houses the reactor is currently a Department of Veterans Affairs Medical Center and will continue to function in this capacity post decommissioning. The planned use of the current reactor areas will be for storage, laboratory space, or staff offices. Based on the site objective, safe storage (SAFSTOR) and entombment (ENTOMB) decommissioning alternatives were considered and rejected as inconsistent with the planned future use.

Decontamination (DECON) is the decommissioning option that has been selected by Omaha VAMC to best support site requirements.

No impact to the surrounding community and natural resources is anticipated to result from the DECON alternative. The Omaha VAMC's current functions and the existing occupancy will not change as a result of decommissioning activities or outcomes. There exists no potential for criticality during any of the DECON activities as all fissile material has been removed from the site.

6.2 Alternatives Considered SAFSTOR or ENTOMB were considered to satisfy requirements for public protection while minimizing initial commitments of time, money, radiation exposure, and waste disposal capacity. However, both SAFSTOR or ENTOMB strategies require prolonged decommissioning schedules and would result in the current reactor spaces being unavailable for extended periods. In addition, maintenance, security, and surveillance would be continuously required until the final decontamination activities were completed. Therefore, both of these approaches would require continued annual expenditures for operating costs and management oversight. Furthermore, over an increased timeline, the availability of personnel with deep knowledge and expertise in reactor operations would diminish significantly. For these reasons, both SAFSTOR and ENTOMB were rejected as decommissioning alternatives.

50

6.3 Rationale for Chosen Alternative DECON is the most environmentally preferable alternative and allows for the maximum flexibility in future re-use of the site. Conformity to ALARA principles will minimize radioactive exposures to the Omaha VAMC staff and patients, as well as contractors participating in the DECON process. This process will be followed by a comprehensive final contamination survey demonstrating that the reactor site meets the NRC criteria for release to unrestricted use. The survey results will be documented and submitted to the NRC to support the request for reactor license termination and release to unrestricted use.

51

7 ALARA Analysis 7.1 Description of How the Licensee Will Achieve a Decommissioning Goal Below the Dose Limit The NRC requirement for termination, 10 CFR 20.1402, Subpart E, provides radiological criteria governing release of a site for unrestricted use. According to this requirement, a site will be considered acceptable for unrestricted use, (1) if the residual radioactivity measures a maximum Total Effective Dose Equivalent (TEDE) to an average member of the critical group (per 10 CFR 20.1003, critical group means the group of individuals reasonably expected to receive the greatest exposure to residual radioactivity for any applicable set of circumstances) that does not exceed 25 millirem (mrem) per year, including that from groundwater sources of drinking water and (2) the residual radioactivity has been reduced to levels that are ALARA. The Final Status survey will use the DCGLs developed from the site characterization report data (Attachment A) and the current NRC guidance for license termination.

The current NRC guidance for acceptable license termination screening values (meeting the 10 CFR 20.1402 criteria) of common radionuclides for building surface and surface soil contamination is presented in Appendix D of NUREG-1727, ALARA ANALYSES. Remediation activities within the AJBRF will be conducted to reduce the residual radioactivity distinguishable from background in accordance to 10 CFR 20.1402, Subpart E so that the site is released for unrestrictive use. Removable surface contamination will be eliminated, where possible, by manual methods. A loose contamination limit of 1000 dpm/1 00 cm2 and a fixed contamination limit of 5000 dpm/100 cm2 shall govern the free release of materials. The DCGL for 25 mrem/yr is 27,000 dpm/100 cm2, as determined by the site characterization report (Attachment A). The calculated DCGL generally exceeds the contamination levels currently present on the structures and surfaces of equipment. Plant systems and reactor components, constitute an exception, and will receive decommissioning to meet the calculated DCGL. Per the site characterization report results, the structures and components within the AJBRF contain non-detectable levels of contamination.

Reactor systems and reactor components, and certain laboratory ventilation hood areas and floor drains will be decontaminated and removed to meet the calculated DCGL. Per the site characterization report results, most of the structures and components within the AJBRF contain very low levels of contamination.

The Omaha VAMC intends to dismantle these structures and components to create usable research space. Correspondingly, most of the internal structures 52

and components will be removed to facilitate the desired future use. Materials, structures, and components that have fixed contamination above 5000 dpm/1 00 cm2 will be dismantled, removed, and disposed off at licensed facilities. Setting a lower limit of 5000 dpm/1 00 cm2 is in accordance with ALARA principles.

Contamination and radiation levels of the reactor internal components and the reactor pit have contamination and/or radiation levels that will require these materials to be handled as radioactive waste (as well as the demineralizer and the resin). Of the remaining areas and components that have been determined to have contamination levels in excess of the calculated DCGL of 27,000 dpm/100 cm2, only the laboratory hood, hood drain, and one sink drain have higher contamination levels. However, in order to accommodate the planned post-decontamination use of the AJBRF, the existing structures and components must be removed. Therefore, the cost of achieving a release limit of 5000 dpm/100 cm2 is only incrementally higher than utilizing a release limit of 27,000 dpml1 00 cm2. The benefit, in terms of the dose to the critical population (the Omaha VAMC staff occupying the spaces), will be exposure to significantly lower levels of radiation.

Surface soil, groundwater and environment have no residual radioactivity distinguishable from background and therefore do not require the demonstration that these levels are ALARA.

7.2 Cost Benefit Analysis Since the planned decommissioning actions will achieve ALARA public dose, a quantitative cost benefit analysis is not relevant to the AJBRF. Section 14 provides additional detail on the rationale and efforts to achieve final levels of contamination well below that calculated to result in 25 mr/yr to the public.

Based on the small number of localized contaminated components in the AJBRF, the resulting work presents a limited risk to decommissioning workers and to the Omaha VAMC population overall during the decommissioning. Finally, factoring the potential benefit of creating usable research, office space in a congested facility, decontamination and decommissioning the AJBRF to allow use as hospital staff office and laboratory space is considered highly beneficial.

The cost of achieving the lower limits of contamination are insignificantly greater than decommissioning to the calculated DCGL level of 27,000 dpm/100 cm2, in that, such lower levels can be achieved by a combination of good housekeeping and the removal of structures and components associated with the AJBRF operation and research activities. The cost of properly shipping and disposal of contaminated materials that exceed would be incurred regardless of the use of the calculated DCGL or the lower release limits. Similarly, the additional cost of controls on the demolition, handling and removal of structures and components to achieve the desired post-decommissioning usability of the facility will be small compared to the overall cost of the project.

53

8 Planned Decommissioning Activities The DVA intends to decontaminate dismantle the AJB reactor and associated systems in a safe manner, in accordance with ALARA principles, the AJBRF Radiation Protection Program, and written procedures. The objective of the AJBRF decommissioning activities is the release of the AJBRF for unrestricted use by the Omaha VAMC. The planned decommissioning activities will include the following general activities:

• Removal of all contaminated equipment and materials, including the reactor internal components and any radioactive or contaminated portions of the reactor tank, and adjacent soil;

• Decontamination of the facility external walls, floors and ventilation ducting as required;

• Removal of decontaminated or non-contaminated equipment, internal walls, and laboratory furnishings;

• Back-filling the reactor pit with clean fill to be covered by a concrete floor after the license termination survey activities are complete;

• Performance of a final radiological survey in accordance with NUREG-1575, MARSSIM to ensure that the internal and external areas meet the criteria for unrestricted release (see Section 14 for the Final Survey Plan);

• Performance of an independent confirmatory radiological survey by the USNRC to ensure the unrestricted release criteria has been met; and

• Placement of concrete pad over the back-filled reactor tank pit and re-modeling of the area as required for use by the Omaha VAMC staff.

The decontamination, dismantling, demolition of the facility, removal of materials (both contaminated and non-contaminated), and reconstruction of the area will be performed by contractors under the supervision of Omaha VAMC staff and independent oversight consultants. Section 9 provides organizational information governing the performance of the effort.

Figure 8.1 is the preliminary decommissioning project schedule presented in Gantt chart format showing the preparatory decommissioning activities, including final survey and license termination. In general, the decommissioning activities will commence with the components and/or structures that are the most contaminated and proceed to those structures and components that are least contaminated. This will minimize the potential spread of contamination to areas considered clean and will remove sources of exposure from the job site.

Therefore, reducing the overall exposure to the decommissioning work crews.

54

Figure 8.1 Preliminary Decommissioning Project Schedule (Gantt chart) rask N3me 6

7l2 Qtr3 8 Mobilization i

M Estabish perlmder secaly and contrd oxales tor access ramp constnxlion Mobilation of personnel ard eqipert El Construction of access to Reactor end establshmert ot P cortrd pocits Establish Radwasteaezadcus Waste Stagng Area atN. Door to Facty E Decontarmination and Dismantlement Actvities E Console, Ventilation, and WalI-In CoolerPit Remediation Decorteminsle and remove matere, eqiprertfrom waek4n cooe r room (SW.2D) and -oa g pit) istal temrrorary exhst vertiation and seal hstsled exhaust ducts Remove reactor console S Reactor wessel, pit rmedadtlon Remove reactor control rod drives and equipment above reactor instal confnemert barrier and temporary steklrig around pod.

Remove accessible Herral reactor components (control rods, igits, dummy elements, instrunertation).

Disassemble, sze and remove retetledor, PTS tbues, core supcrt dtrudcae U

Establish cortroled wea ad stage water processing/Storage erIpment at demintieat exchanger pt instal temporary connedionspiping to reactor water heat exchanger Remove reactorpod water,translerto storagetank fordisposair.

Removel of concrete and gurle on D of reactor vessel 55

Figure 8.1 Preliminary Decommissioning Project Schedule (Gantt chart) (con't) akae M171IM¶8'M191

-- Tm-- i 2

Decontamn*ate reactor vessel as necessary Perform test bores in reactor vessel, deterirne exterd of conitaminated, irradiated material, concretelsoil in reactor pi M~sal P1 temporary brackVg, section andl remove reactor vessellf Sample. stzvey reactor pi corcrete wall Remove cortamirnated reactor p1 concrete and surrounding adt as necessary Post remediation survey of reactor PI Cover p1 wth steel plate and remove confinement berrier B Heat Exchangeilon Exchange System Decomrmlssionlng Remove and flush resion for Ion excharger Flush. drain, dissembtle-and decontaminate heat exchargerAon exchargeir system Decortatriiate and seed penratisons from pi to reactor faciliy Decortarriale and remove contamirnated materials, components from Heat Exchang Pt Survey pt, close access, cleanup~ and remove area cortrols 8 Decommission labs and structure.

DeconfIremove lab snk~scabinets; decondkahs ~Inrooms SW-2C and SW-2E Decen Vaut room (SW2I') vauts, wetls Decon vault room drains Decon Vault Room Floor 8 SVN-2E Decon Main Laboratory Room (SN-2) drans* and storage pis Decon as necessary, remove Inerior walls of rooms SW-2C and SW-2 Decon and remove as necessary wall, cefirg, gmoo materials from rooms S~N-213 and SW.2A3

'ask Name Decon andl remove as necessary wet, cedirg. floor materiIal from SW-2 General oitslde cleanup.

Rad Waste Pick up for Transportation8 Disposal embiation E3finaeI S-urve-ys end D-W0oe um -e ratstion Finl.urvysandDoumetaio Prepare, R&A Final Survey report Subitrr Final Survey report to NR~C NRC verification survey 8 sle release an eolnof ftn~ns Projct Closeout; Transfer o1 records. drawirgs, fles, and documjents to OVA.

Bactill reactor p1 area, pour concrete footerlfloor I

M I

a

-Y UM 1

Seal temporary access to faciltly, backttl ramp~,

p1 areas___

56

Pre-Decommissioning Plan Approval Activities The planning, preparation and mobilization activities shall ensure that the necessary resources, procedures, work requirements, controls and processes, Omaha VAMC and contractor roles and responsibilities are clearly identified and documented.

General Cleanup As part of decommissioning planning actions, a site characterization report was conducted by GTS Duratek to determine the type, quantity, condition and location of radioactive and/or hazardous materials present in the Omaha VAMC and surrounding areas. This comprehensive radiological survey was completed in December 2002 and the data and results of this survey are included in Attachment A to this plan.

Additionally, a general cleanup of the AJBRF was conducted to identify, properly package and dispose of chemical and radiological wastes and non-reactor related equipment and materials. This cleanup was conducted in December 2002 and included disposal of the licensed sources as described in Table 2.1, except for the Am-Be neutron source, the disposal of which is pending acceptance by Los Alamos National Laboratory's Orphan Source Recovery (OSR) program.

The office furniture and equipment, documents, books and tools will be surveyed for free release, disposed of or re-assigned for use at the Omaha VAMC, and the remaining non-radioactive/contaminated materials removed from the AJBRF prior to decommissioning mobilization.

The floor tiles inside the AJBRF were determined during the site characterization to contain asbestos. Those floor tiles found not to be radiologically contaminated will be removed by a qualified asbestos abatement vendor during the pre-decommissioning planning phase. This work will be done in accordance with the Omaha VAMC safety manual and 29 CFR.1926.1 101.

Those uncontaminated wall areas painted with lead-based paint will also be removed prior to the start of the actual decommissioning work. This work will be performed by a qualified contractor in accordance with the Omaha VAMC Safety Manual and the requirements of 29 CFR 1926.62.

The existence of the asbestos and lead materials present the primary safety and non-radiological remediation issues associated with the decommissioning of the AJBRF, beyond the industrial and radiological safety issues expected in a research reactor decommissioning. As noted above, all the chemicals and 57

chemical waste that were generated or stored in the AJBRF laboratories have been removed and disposed of properly.

Decommissioning Engineering During the decommissioning planning phase the project management team, the Quality Oversight Team, and Safety Program Manager will direct and implement the Omaha VAMC Quality Assurance (QA), safety and radiological protection programs. These specifications will be incorporated into the decommissioning contractor bid specifications. The bid specifications shall include decontamination, demolition, radwaste packaging and disposal, and final status survey activities.

The decommissioning engineering contractor will be assigned preliminary engineering tasks to support the decommissioning work. Preliminary engineering tasks will include the design of temporary electrical power, temporary exhaust ventilation, temporary structural support for the access ramp, as well as the planning for the removal of reactor systems and reactor components.

In conjunction with the preliminary engineering tasks, the technical specifications shall be reviewed to determine which, if any, required surveillances will be impacted by the decommissioning activities. Requests for changes to the technical specifications will be prepared and submitted to address such impacts prior to the start of the actual decommissioning work.

Decommissioning Preparatory Activities

1. Procedure and Work Control System Development Prior to mobilization for the start of the decontamination and demolition activities, the decommissioning contractors', the safety and oversight team, and the Omaha VAMC decommissioning personnel (i.e., RSO and Safety Manager) shall prepare procedures for project planning, control and oversight, and reporting of work.

The RSC will review all procedures for consistency and completeness, as well as compliance with NRC, the State of Nebraska, and Omaha VAMC requirements and regulations, as well as the AJBRF license. Oversight and QA procedures will be developed to ensure that work is performed in accordance with written procedures and to ensure that the safety and radiation protection measures are properly controlled are implemented.

58

2. Site Radiological Survey Prior to mobilization, a confirmatory radiological survey will be conducted and compared to findings of the previously performed site radiological characterization. Additionally, the survey activities will include smears to determine the facility contamination levels, general radiation exposure rates and location of any hot spots.

During the decommissioning project, additional surveys will be conducted in accordance with Omaha VAMC radiation protection procedures and Section 10 of this decommissioning plan to ensure proper work practices and protective equipment and clothing are used.

3. Site Orientation/Radiation Protection TraininglRadiological Screening Each individual performing decommissioning work, oversight or supervision shall attend a one-day site-specific Radiation Worker training in accordance with approved Omaha VAMC procedures and 10 CFR 19.12. Additionally, all workers shall attend a four-hour class on Omaha VAMC safety policy and procedures. All personnel whose work will require entry into the controlled access areas (CAAs) will have initial and exit bioassays performed.

Mobilization Mobilization will occur in stages as required by the project plan (Figure 8.1).

Initial mobilization of personnel will include the travel of workers to the site and acquisition and delivery of required equipment for the tasks listed below. In addition to the Oversight and Safety Team and the Decommissioning Project Management team, the initial mobilization shall consist of the contracted supplemental radiation protection personnel required to support the performance of the following tasks.

1. Establishment of Alternate Access to AJBRF To minimize disruption of hospital operations during decommissioning and to minimize the potential for delays in accessing the reactor area and delays in removing contaminated materials, an alternate access shall be constructed from outside the Omaha VAMC research building (Figure 2.1) to the AJBRF. This access will consist of a ramp dug through the storage pit exterior and through a 59

temporary door in the exterior or west wall of the walk-in cooler (room SW 2D in Figure 2.1).

2. Establishment of Radiological and Hazardous Waste Packaging/Staging Area The area immediately inside the north door to the facility, previously the normal access route, will be utilized to package industrial, radiological, and hazardous waste. Once packaged, the waste will be taken via the alternate access ramp and staged in the area immediately outside the control point for loading onto transport vehicles.

3. Establishment Exterior Controlled Area Boundary and Radiological Control Points The areas outside the facility will serve as an equipment staging area (Figure 2.1). These areas will be used for placement of temporary exhaust ventilation fans and high efficiency particulate air (HEPA) filters and for staging a trailers for the storage and donning of protection clothing, offices for supervisory personnel, break rooms for workers and the staging of packaged radwaste for pickup. A step-off pad will be established at the bottom of the access ramp and at each of the pre-existing access points at the north and south of the facility, as necessary.

Each step off pad area will have bins for collecting used protective clothing and trash, frisking stations and space for logging in and out of Radiological and Hazardous Work Permits (RHWPs).

The site boundary will be extended to include the ramp and a radiological control point will be set up at the entrance to room SW 2D. The entire exterior area to the west of the building housing the AJ8RF will be a controlled area and attended by radiation protection personnel during working hours. During non-business hours, the door to the temporary gate to the reactor facility will be locked.

The area east of the facility, in the vicinity of the reactor cooling system will be where the primary HEPA exhaust ventilation system will be staged, utilizing the existing exhaust fan and/or opening. Additional HEPA systems to exhaust from the facility will be utilized as needed, and may be staged either to the east or west (i.e. near the access ramp).

During the D&D effort, the dedicated force of Omaha VAMC police will provide security. They regularly patrol the hospital grounds and interior. Also, a closed circuit camera will be mounted to provide 24-hour surveillance utilizing the Omaha VAMC police closed circuit television system.

60

4. Radiological and Industrial Safety Controls All work performed within the ABJRF will be conducted in accordance with the AJBRF Radiation Protection Program and approved radiological control procedures (as summarized in Section 10 of this plan). All personnel assigned to work in the AJBRF will be qualified as 'radiation workers" requiring internal and external radiation exposure monitoring, medical qualification for wearing respiratory protection, and trained in accordance with AJBRF radiation protection procedures and 10 CFR 19.12.

All work performed in the AJBRF shall be done under a RHWP reviewed and approved by the Radiation Safety Officer (RSO) and Safety Program Manager.

Each task performed will be evaluated for the use of personal protective equipment and radiological controls.

The minimum personal protective equipment (PPE) required shall be eye protection, hard hats, and work shoes. Hearing protection will be required for activities that produce noise levels exceeding 85 decibel (dB) in accordance with 29 CFR 1910.95. Face shields and dust protection shall be worn during cutting, grinding, scabbling or demolition of non-contaminated structures.

Work on contaminated systems shall require all workers to wear, as a minimum, coveralls, shoe covers, and rubber gloves, in addition to the required PPE above.

Respirators will be worn at the discretion of the RSO and, if the wearing of such equipment is required to reduce exposure, during cutting, grinding and/or scabbling of contaminated equipment and structures.

Additionally, dedicated ventilation systems with HEPA filters shall be used for cutting into contaminated systems and during work inside the reactor pit. The use of containment structures and glove boxes shall be used to minimize the potential for airborne contamination.

Additionally, Radiation Protection Technicians, directed by the RSO, shall monitor work and working conditions to ensure proper controls and protective equipment are utilized during each activity.

Table 8.1 lists the expected radiological and industrial safety hazards and mitigating measures to be employed to limit risks to the workers, Omaha VAMC staff and patients, and the public.

61

Table 8.1, Expected radiological and industrial safety hazards and mitigating measures Radiologicallndustrial Safety Risk

Mitigating Measures ii>>,

Radiological Risks High radiation exposure during:

Reactor internal components removal Use of remotely operated underwater tools; Mock-up training; and Use of specialized shielding intermediate casks for internals removal from core and transfer to shipping cask.

• Reactor tank/pit materials removal Task planning using ALARA principles; Use of specialized shielding/shielded work platform; and Use of remotely operated tools.

Airbome contamination during:

Reactor internals removal Worker training on and use of respiratory Reactor tank/pit materials protection equipment; decontamination and removal Work area containment enclosures; Reactor cooling, purification system and Portable vacuum HEPA filtration and ventilation pneumatic transfer system piping and systems; components removal Pre-removal decontamination to eliminate loose Facility drains, hood and structures surface contamination; and decontamination and removal Worker use of proper protective clothing.

Industrial Risks Confined spaces hazards during:

Reactor Tank/Pit remediation Confined space entry and controls training; Cooling Pit equipment removal and area Use of tenders for each worker; and remediation Sampling of atmosphere at start and during work.

62

Table 8.1, Expected radiological and industrial safety hazards and mitigating measures (con't)

Radiological/industrial Safety Risk Mitigating Measures Electrical Safety Hazards risk during:

Removal/movement of facility electrical Worker training on electrical safety and lockout/

power supplies tag out procedures; Demolition/removal of De-energize AJBRF electrical systems (except structure/components adjacent to lights and local receptacles); and electrical systems Use of ground fault circuit interrupters.

Excavation stability hazards during:

Access ramp construction

• Licensed engineered design for access ramp and temporary supports; On-site technical direction by responsible engineer; Excavation contractor training of workers on excavation safety; and Use of sloping sides/benching for access ramp excavation.

Reactor pit remediation Installation of engineered, temporary bracing in reactor pit; Responsible engineer, safety manager inspection of installed bracing prior to start of work; Contractor training of workers on excavation safety; and Daily inspection of bracing, pit walls and after removal of sections of tank/concrete pit wall material removals.

63

Table 8.1, Expected radiological and industrial safety hazards and mitigating measures (con't)

Radiological/Industrial Safety Risk' Mitigating Measures E Welding, cutting, burning, hot work risks during:

• Removal of equipment, components and system & drain piping

• Removal reactor tank/pit structure

• Removal of facility walls, laboratory structures

• Installation and demolition of temporary supports, bracing

• Worker training of hot work safety precautions and procedure;

• Surface decontamination of surfaces to be cut, welded;

• Asbestos and lead paint abatement prior to cuffing, welding;

• Use of hot work permits per Omaha VAMC procedure;

• Assigned fire watch for each cutting, welding task; and

• Daily housekeeping to minimize combustible material accumulation.

+

Personnel Fall risk:

• Installation of temporary bracing in reactor pit, access ramp

• Remediation of reactor pit

• Lead paint abatement Demolition of laboratory walls, structures

• Worker training on fall protection and use of scaffolding;

• Pre-job inspection of scaffolding prior to use;

• Erection of scaffolds by qualified personnel;

• Use of fall protection equipment - handrails, toe boards and covers over openings;

• Provision of protection from falling objects, i.e., canopies, cordoning of areas below work areas; and

• Use of personnel fall arrest systems.

64

Table 8.1, Expected radiological and industrial safety hazards and mitigating measures (con't)

Radiologicallndustrial Safety Risk l Mitigating Measures 1;:->-

Welding, cutting, buming, hot work risks during:

• Removal of equipment, components and system & drain piping

• Removal reactor tank/pit structure

• Removal of facility walls, laboratory structures

• Installation and demolition of temporary supports, bracing

• Worker training of hot work safety precautions and procedure;

• Surface decontamination of surfaces to be cut, welded; Asbestos and lead paint abatement prior to cutting, welding;

• Use of hot work permits per Omaha VAMC procedure;

• Assigned fire watch for each cutting, welding task; and

• Daily housekeeping to minimize combustible material accumulation.

Mobile equipment and material handling during:

• Worker training in Contractors' safety

• Installation of special tooling, procedure for safe handling of materials and equipment over reactor pit mobile equipment; and

• Removal of equipment and packaged Worker training of handling and storage of waste from facility radioactive waste.

• Placement of equipment and waste on trucks for transport Power tool use during:

Mobilization, decontamination and Contractor training of workers in proper use dismantling tasks and maintenance of power tools; Omaha VAMC oversight to ensure tools are in good condition and incorporate safety features; and Contractor supervision and Omaha VAMC oversight on proper use during performance of tasks.

65

8.1 Contaminated Structures Contaminated and potentially contaminated AJBRF structures consist of 1) those items that may have been exposed to neutron activation including materials composed of aluminum, steel, stainless-steel, graphite, and concrete, and 2) those areas of the AJBRF walls, floors, ceilings, and drains that may have been contaminated with radioactive material during the AJBRF operations, (maintenance as well as during sample preparation and handling).

Structures analyzed as activated consist of the reactor tank, surrounding concrete and immediately surrounding soil. Referring to Figure 2.1, the areas found to have contaminated structures, were rooms SW 2, SW 2C SW 2D, and SW 2F, and the heat/ion exchanger pit outside of the east wall of room SW 2A.

The walls, floors and ceilings of rooms SW 2A and SW 2B were found to have contamination levels well below the limits of 5000 dpm/1 OOcm 2 for fixed contamination and 1000 dpm/100cm 2 for removable contamination. This correlates with the historical use of these rooms for research utilizing only short-lived isotopes until the early 1990s and subsequent use for activities involved in cancer research that did not utilize any radioactive materials. Additionally, the areas immediately outside the AJBRF site boundary (i.e., SW 1A and the hallway and storage room to the north of the facility, as well as the stairwell at the south end of the facility) showed no evidence of contamination greater than 1,000 dpm/1 00cm2.

The overhead ceiling tiles and supports were found to be uncontaminated. After the initial pre-decommissioning radiological survey, the ceiling tiles and support structures will be removed. The lighting fixtures and ventilation ductwork will remain and be replaced or modified after the decommissioning effort is completed.

Specifically, the contaminated structures at the AJBRF include the following:

8.1.1 Reactor The reactor is located at the bottom of a cylindrical pit 6.1 m below the floor level in room SW 2. The reactor structure consists of a carbon steel tank 0.64 cm thick with an inner diameter of 208 cm resting on 28 cm concrete slab.

Approximately 25 cm of poured concrete surrounds the outside of the tank. The inside of the tank is covered on the sides by a five cm thick layer of gunite and on the bottom by ten cm of poured concrete. The entire inner surface is coated with two applications of a waterproof epoxy resin coating.

66

The center channel assembly, control rod drive mechanisms and aluminum grating/plexiglas tank covers will be manually decontaminated to the extent possible, removed and cut to size for transport to a decontamination/ disposal site. Subsequently, the core internal components, including the reflector, the upper and lower support plates and the guide tubes and control rods, as well as remaining lights, instruments and sections of the pneumatic transfer system will be removed prior to any decontamination and dismantlement of the reactor structure (see Section 8.2.1 below).

A containment enclosure will be constructed around and over the reactor pit, with a dedicated ventilation and HEPA filter system to prevent the spread of contamination into the rest of the facility. Disassembly and/or cutting of the core components, as described in Section 8.2.1 below, will be performed under water to minimize exposure to the workers.

Upon removal of the reactor internal components and water (as described in Sections 8.2.1 and 8.2.2) core bores will be performed into the surrounding material to determine the depth of irradiated material, i.e., epoxy/gunite, steel, concrete and surrounding soil. Depending upon the extent and levels of contamination/ activation, some or all of the tank/pit structure may be left intact.

After the termination survey, the pit would be backfilled and sealed with a concrete slab. If contamination/ activation levels are such that removal of some or all of the pit structure is required, then the tank will be cut in horizontal sections, starting from the top down. After removal of the steel tank, the support bracing will be installed inside the pit and the concrete and soil determined to be irradiated will be removed.

The Omaha VAMC research wing, including the AJBRF, sits on 1.5 m thick concrete footers that effectively surround the reactor pit, approximately 46 cm from the outer diameter of the reactor tank. To ensure that the pit does not collapse upon itself during the decommissioning, engineered supports will be installed to support the sides of the pit. The reactor pit will be classified as a confined space requiring continuous air monitoring, lanyards on workers and tenders for each worker in the space. Each individual who works in the reactor pit shall be trained for confined space entry. All work in the reactor pit will be in accordance with written procedures that incorporate radiological and industrial safety controls that have been approved by the RSC.

Upon completion of the reactor pit decommissioning, the pit will be sealed with plastic sheets and a steel plate will be placed on top of pit. The engineering supports and bracing inside pit will remain in place until the pit is backfilled and a concrete floor poured over it.

67

8.1.2 Room SW 2 Three emergency fuel element storage pits are located immediately adjacent to the reactor tank. The pits are vertical steel pipes 25 cm in diameter and 305 cm long, and are lined with an organic coating. No evidence of fixed or removable contamination was discovered in these storage pits. Based on the pre-decommissioning site survey results, these pits will be cleaned of any debris, decontaminated, as required, and sealed. After the site termination survey, these pits will be filled with concrete.

The trench that runs along the exterior of rooms SW 2A and 2B and diagonally to the reactor pit contains the pneumatic transfer system tubing. The pneumatic transfer system tubing and miscellaneous cables, and loose material will be removed, sorted and shipped as radwaste as required. The trench surface is concrete, which will be scabbled to remove contaminated materials. The removal of the contaminated concrete will be performed in conjunction with the removal of the floor tiles and any contaminated concrete surfaces below the tile.

The only area of general contamination of the floor of room SW 2 occurs in the area between the reactor console and the reactor pit. This area will be decontaminated by removal of the floor tiles and adhesive, followed by decontamination of the underlying concrete by scabbling.

8.1.3 Room SW 2F Room SW 2F is a vault room used primarily for storage of radioactive sources and radioactive waste. The south wall contains 64 vaults, each approximately 25 cm deep. Additionally, the floor of this room contains storage pits previously used for storage of sources. The floor, south wall and vaults are contaminated.

The vaults and pits have steel liners that will be manually decontaminated in place and then removed by cutting. Scabbling or grinding out the inner concrete surfaces will remove any contamination remaining after liner removal. Initially, the floor of room SW 2F will be decontaminated manually and, if necessary, the concrete will be removed by scabbling.

8.1.4 Rooms SW 2C and SW 2D The contaminated structures in these two rooms are the walls of room SW 2C.

These walls will be decontaminated by a combination of manual methods and subsequent removal of contaminated material by cutting out of the contaminated areas. Additionally, the walk-in cooler floor and wall (room SW 2D) show relatively higher contamination than the other structures, although the maximum level is below 1000 dpm/100 cm2. Therefore, the floors and wall of the walk in cooler will be manually decontaminated and re-surveyed. Additional decontamination requirements and approach will be determined at that time.

68

8.1.5 Heat and Ion Exchanger Pit The Heat and Ion Exchanger Pit (referred to as the East Cooling Pit in the site characterization report, Attachment A) floor is contaminated in the area of the ion exchanger. This area will be decontaminated manually, with removal of any subsequent fixed contamination by removal of the paint and surface concrete.

The decontamination of the structures (walls and floor) will occur after the removal of the piping and components, and sealing of the piping penetrations through the east wall of the facility.

8.1.6 Asbestos and Lead Paint Abatement The walls that make up the AJBRF site boundary are painted with lead-based paint. This paint will be removed prior to the commencement of the decommissioning activities on all wall areas that are not contaminated. Any wall area with lead-based paint that is contaminated in excess of 1000 dpm/1 00cm2 will have those areas removed during the decommissioning and prior to dismantlement of the associated walls. Lead paint removal will be done in accordance with the Omaha VAMC safety manual and 29 CFR 1926.62, and in such a manner to minimize the release of material into the atmosphere. Only individuals who have been trained in lead abatement will be utilized for this task.

The materials removed will be handled and disposed of as hazardous mixed waste, or waste, as is legally appropriate.

Asbestos exists in the floor tiles and floor tile mastic adhesive. The levels of asbestos in the floor tiles and adhesive range from approximately 10 - 25 weight%. Removal of the non-radioactively contaminated tiles and adhesive will be performed as part of the actual pre-decommissioning activities above. The floor tiles that have been shown to be contaminated have low levels of removable activity, and will be decontaminated manually to minimize damage to the floor tiles and potential release of asbestos into the air. Qualified and experienced vendors requiring respiratory protection and a dedicated ventilation system will perform the asbestos removal in compliance with the Omaha VAMC safety manual and 29 CFR 1926.1101 8.1.7 General Clean up and Removal of Uncontaminated Structures The walls of rooms SW 2E and SW 2C will be removed to enlarge SW 2 during the decommissioning. Walls that have fixed contamination will have the contaminated areas removed as described in 8.1.6 above. Once the walls have been decontaminated, they will be demolished and removed from the facility.

Removal of interior walls will occur after any asbestos/lead abatement efforts and decontamination/removal of adjacent equipment, systems and/or materials.

69

8.2 Contaminated Systems The decontamination, dismantlement and removal of the various reactor systems will be performed in the sequence shown in the proposed schedule provided in Section 8.1.

8.2.1 Reactor Internal Components The reactor internal components consist of:

Reflector-an aluminum cylinder 30.5 cm thick, 43 cm ID, 107 cm Outer Diameter (OD); and 56 cm high, and containing aluminum cans that contain graphite.

• Top and Bottom Grid Plates - aluminum grids approximately two cm thick that support and align the fuel elements.

• Rotary Specimen Rack - an aluminum ring that contains forty evenly spaced aluminum cups that are hung from the ring and served as irradiation specimen holders. The rotary specimen rack is completely enclosed in a welded aluminum box.

• Pneumatic Transfer Tube and Central Thimble - aluminum tubs that either enter from the side of the tank or vertically from the top. They were used to send samples into, and retrieve from, the core.

• Control Rods and Guide Tubes - Three boron carbide control rods that operate inside aluminum guide tubes for control of the reactor power.

Aluminum extension tubes connect the control rods to the control rod drives on top of the reactor.

• Lights and Visual Inspection Tool - Aluminum fixtures used in operation and maintenance of the reactor.

• Dummy Fuel Elements - Aluminum-clad elements in which the uranium-zirconium-hydride fuel has been replaced by graphite.

These components will be cut and/or disassembled underwater by experienced vendors utilizing remote tooling. This will prevent generation of airborne contamination and provide shielding for the workers. After cutting/disassembly, the pieces shall be raised to the surface and placed inside shielded containers for movement to the appropriate shipping containers. These containers will be loaded, drained of water and the internal components shall be shipped to decontamination/disposal sites.

During this evolution the reactor water-cooling and demineralizer system will remain in operation to maintain water clarity and remove contaminated particles generated during the operation. Continuous air monitoring will be performed and all workers will utilize appropriate protection clothing and respiratory protection 70

equipment. Additionally, all tools and materials being removed from the core will be surveyed underwater prior to removal to minimize the potential for excessive radiation exposure and to detect any hot particles.

8.2.2 Reactor Cooling and Purification System The reactor cooling and purification system consists of a five-ton freon vapor-compression chiller with an air-cooled condenser, water filter and mixed-bed demineralizer, instrumentation and piping. The chiller and demineralizer are located in the East Cooling Pit, outside the Omaha VAMC research building.

Piping between the reactor and the demineralizer and chiller penetrate the east wall of AJBRF room SW 2 approximately 3.65 m above the floor of room SW 2.

After the reactor internal components are removed, the reactor coolant water will be pumped from the reactor, through an ion exchanger and into a tank staged outside the temporary access. The water will be sampled, stabilized, and sent to a disposal site.

The system piping will be cut out, with precautions taken to prevent pinching the ends closed to allow decontamination of the inner surfaces. All cutting operations will be performed in conjunction with the use of dedicated exhaust ventilation that passes through HEPA filters and with the use of respiratory protection, as required. The demineralizer resin will be discharged and transported to a disposal site for burial. The demineralizer shell shall be surveyed and sent to a decontamination/disposal site.

The chiller will be drained into appropriate containers opened up and decontaminated, if necessary. Based upon the very low radiation levels associated with this equipment, very little contamination is anticipated from this component. The freon will be discharged into licensed container and transported by an authorized contractor for disposal.

The penetrations in the east wall of room SW 2 will be surveyed, decontaminated by manual methods as necessary, and sealed.

8.2.3 Pneumatic Transfer System The pneumatic transfer system tubing that exists outside the reactor pit will be removed and/or cut, surveyed, and decontaminated. The tubing inside the walls will be sealed inside the walls/floors if it is not contaminated. If the inner surfaces of the pneumatic transfer system aluminum tubing cannot be decontaminated, then the floor/wall materials will be removed and as will be the tubing.

Precautions will be taken to avoid pinching closed this tubing during cutting and all cutting will be done in conjunction with a dedicated exhaust ventilation system and HEPA filters to prevent spread of airborne contamination. Personnel respiratory protection will be utilized as necessary. All removed tubing will be 71

transported and disposed off of-site. The pneumatic transfer system trenches in the AJBRF will be remediated as indicated in Section 8.1.2 above.

8.2.4 Facility Ventilation System No evidence of contamination was found in the ventilation system ductworks that lead from the reactor room to the roof of the hospital. Therefore, the exhaust ventilation system will be sealed at the start of the decommissioning effort to prevent contamination of this system due to the decontamination and dismantlement activities.

The supply side of the ventilation system will remain active, and due to the positive air pressure in the system relative to the reactor rooms, it is not anticipated that the supply train will become contaminated. The use of a temporary exhaust ventilation system that exhausts through HEPA filters to the outside environment will also prevent contamination of the installed ventilation system. Therefore, no remediation of the ventilation system is anticipated.

8.2.5 Radiation Monitoring System The radiation monitoring system at the AJBRF consists of two continuous air monitors (CAMs) and a gamma monitor with an alarm set at two mrem/hr.

Similar instruments will remain in operation during the decommissioning. If replacement or supplemental instrumentation is required for the decommissioning process, similar instruments with like capabilities will be utilized.

8.3 Soil and Exterior Areas Surveys of the exterior surroundings and surface soil during the site characterization found no indication of radioactivity outside the AJBRF. The areas surveyed included the roof of the building that houses the AJBRF and is the exhaust point for the AJBRF ventilation system, as well as the sidewalks and parking areas exterior to the Omaha VAMC research building. These findings correlate with the AJBRF operational history and records that indicate there were no releases or spills of radioactive material during the reactor lifetime. Therefore, no remediation is anticipated for surface soil, sidewalks and parking lot areas at the Omaha VAMC site.

Characterization of the subsurface soil both exterior to the facility and adjacent to the reactor was performed by conducting borings and sampling the removed soil for radioactivity and activation. The geological and radiological analysis, which indicates no detectable radioactivity, is provided in the site characterization report (Attachment A). Therefore, no remediation is anticipated for the subsurface soil outside the reactor facility.

72

The vertical borings inside the AJBRF were located approximately 0.3 m from the edge of the reactor pit. The angular boring retrieved samples below the reactor from 0.6 to 3.1 m below the reactor. None of the soil samples from the borings showed any measurable radioactivity or activation.

During the decommissioning of the reactor tank, horizontal bores will be made approximately at various depths of the pit in each quadrant to determine if the soil immediately outside the tank is activated or contaminated. Any activated or contaminated soil will be removed and shipped to a processing facility for decontamination and/or disposal. Workers utilizing respiratory protection will execute such removal activities, the pit will be sealed by a containment structure, and a dedicated ventilation system will be installed to maintain a negative pressure in and over the pit relative to the reactor facility. Based on the low power of the AJBRF, the low total power produced over its lifetime and the length of time since the last power operations, little activation is anticipated and, subsequently, the radiation levels associated with any potentially activated soil are expected to be very low. The results of the activation analysis performed as part of the site characterization survey supports this conclusion.

8.4 Surfaces and Groundwater The Omaha VAMC is located on the top of a knoll and the nearest surface water to the AJBRF is approximately two miles away (see Section 3.7). Therefore, no remediation of any surface water is anticipated.

No evidence of radioactivity was discovered in any of the samples taken in the borings. Additionally, the depth at which groundwater was found ranged from between 19.2 and 21.9 m below the surface, or approximately 10.7 -12.2 m below the bottom of the reactor pit. Therefore, no remediation is anticipated to be required of ground water under the Omaha VAMC.

Monitoring wells were installed in each of the three vertical borings external to the Omaha VAMC facility (see Appendix D of Attachment A, and Figure 4.2).

These wells will be maintained until the final termination survey is completed.

8.5 Schedule Decommissioning will occur sequentially as detailed in the schedule shown in Figure 8.1. It should be noted that the dates reflected in this schedule are contingent upon NRC approval of the decommissioning plan. Upon approval of the plan a revised schedule will be provided that contains more accurate and firm date estimates. After the initiation of the decommissioning, if it is determined that the decommissioning will not be completed in accordance with the schedule provided, the Omaha VAMC will provide an updated schedule to the NRC.

73

9 Project Management and Organization The DVA is committed to a safe decommissioning effort. The DVA retains ultimate responsibility for the performance of all work in accordance with the AJBRF license requirements. This responsibility includes conformance with the AJBRF Decommissioning Quality Assurance (QA) plan as well as Radiological and Industrial Safety program for all work performed on site.

The decommissioning project management and organization is designed to ensure that the appropriate management control and oversight will be exercised during decommissioning activities to ensure compliance with federal, state and local regulatory requirements and the AJBRF license.

9.1 Decommissioning Management Organization During decommissioning planning, the AJBRF Decommissioning Organization will be formally organized and appropriately staffed. Project governance will be defined prior to project initiation including definition of roles and responsibilities, division of authority, reporting structures, and lines of communication. In addition, project administrative procedures will be prepared and documented in a manner supporting staff training prior to project initiation. The following decommissioning project management and oversight organizations will be established in order to ensure that the decommissioning activities of the facility are completed safely and to the highest standards of quality and performance.

Figure 9.1, Decommissioning Organization, below, details the proposed staffing and lines of reporting.

74

Director, Omaha VAMC Project Controls & Oversight Reactor Facility PlanningISceduling Manager O OT Ucensing lDecommissioning Operations I

Contractor Radiological & Radioactive Waste Decontamination & Dismantlement l

~~~~~~Servces~eotmnbn&Dsnnlmn Operations I

Training I

Radiation Protection Services -

Quality Control L -

personnel and Instrumentation Engineering Radiological Waste Project Scheduling/Cost Final Survey Control Radiological Controls Decontamination Non-Radioactive Waste Dismantlement Disposal Excavation Figure 9.1, Decommissioning Organization Descriptions of key positions in the organization and their responsibilities are outlined below.

9.1.1 Director, Omaha VAMC The Director of the Omaha VAMC has the ultimate responsibility for all functions of the Omaha VAMC including all licensed activity at the AJBRF. The Director shall ensure that adequate funds are available to complete the decommissioning and final radiation survey activities in an efficient and timely fashion. The Director has delegated authority for the overall management and oversight of the decommissioning activities at the AJBRF to the Omaha VAMC Associate Chief of Staff for Research (ACOS/R). This delegated responsibility is to ensure the proper level of management oversight is given to the performance of all decommissioning activities. Additionally, to further ensure proper management oversight of the decommissioning activities, the RSC shall report the results of its audits and meetings to the Director.

75

9.1.2 Associate Chief of Staff for Research The ACOS/R is responsible for ensuring that all decommissioning activities are conducted safely and that the control of radioactive materials and radiation exposure is in accordance with regulatory requirements and ALARA. The ACOS/R shall be responsible for the conduct of the Quality Assurance program and for the effectiveness of the Radiological and Industrial Safety programs. The ACOS/R shall have final approval authority of minor changes to the Decommissioning Plan and to procedures (that do not involve an un-reviewed safety question), and for the daily conduct of the AJBRF decommissioning. The ACOS/R shall also act as the Chairman of the RSC. The ACOS/R shall ensure that:

• Decommissioning activities comply with applicable federal, state and local regulations;

• Decommissioning activities are performed in a manner to protect the public, hospital staff, patients and the environment;

• Proper resources and qualified personnel (or organizations) are assigned adequately to safely achieve the decommissioning of the AJBRF;

• Decommissioning work is performed safely, within budget and on schedule;

• Radiological control and industrial safety program and personnel are effective and have adequate support from management; and

• Contracts, subcontracts, budgets and schedules are approved and awarded in a timely fashion to ensure satisfactory completion of the AJBRF decommissioning.

9.1.3 Project Controls and Oversight 9.1.3.1 Quality Oversight Team (QOT)

The QOT shall ensure that all quality and regulatory requirements are satisfied throughout the entire decommissioning process, from planning, procurement and contracting to decontamination, demolition, waste disposal and final release survey. Project/vendor QA programs and procedures shall control all decommissioning activities. All vendor QA programs and procedures shall be reviewed and approved in conformance with the AJBRF Decommissioning Quality Assurance Program Plan. The QOT shall be responsible for the development, maintenance and execution of the quality controls, policies, practices, and procedures for the decommissioning project with ongoing surveillance and monitoring of key activities and operations. Responsibilities include:

76

• Ensure the contractor's 10 CFR 50 Appendix B QA program is applied to all contractor and sub-contractor activities; Initiate Nonconformance Reports (NCRs)/Corrective Action Reports (CARs) as required and ensure corrective actions are identified and performed;

• Review decontamination and dismantling plans and procedures to ensure proper levels of controls during the execution of the decommissioning activities, and compliance with the requirements set forth in the facility license and applicable regulations;

• Evaluate and survey procedural and RHWP compliance by workers and supervision;

• Review dismantling plans to ensure proper configuration control is maintained to ensure radiological and environmental safety;

• Review receipt and shipment of radioactive materials and equipment such as instruments, sources, or special tools;

• Review all design, procurement, testing, and instrument calibration activities and records; Monitor and conduct surveillance of instrumentation calibrations performed at the Omaha VAMC by RPTs and by approved vendors at vendor facilities;

• Monitor the chain of custody of contamination samples to ensure traceability, control and accuracy of test results;

• Audit vendors who supply health physics instrumentation, equipment and calibration/maintenance services to ensure conformance with federal, state and local regulations and the AJBRF Radiation Protection plan and procedures;

• Audit and approve QA programs of vendors utilized for safety related activities;

• Perform QA audits of the decontamination and decommissioning activities;

• Report regularly to the RSC on the status of quality controls, compliance with procedures, QA, safety, and radiological control plans; and

• Maintain documentation of QA audits and inspections.

The QOT shall develop and/or review procurement and contract documents to ensure that technical, planning and quality specifications are correct and properly detailed. Additionally, post-bid evaluations shall be performed by the QOT.

Decommissioning contractors' personnel qualifications shall be reviewed by the QOT prior to the awarding of contracts.

The QOT team shall report to the ACOS/R. Any member of the QOT shall have the authority and responsibility to interrupt or suspend any activity if it is deemed to be performing in an unsafe, uncontrolled or improper manner.

77

9.1.3.2 Planning/Scheduling Decommissioning planning shall be the responsibility of project controls and oversight team. The decommissioning contractor and the Omaha VAMC project controls personnel shall refine the decommissioning plan and schedule, and the Project Controls and Oversight team shall be responsible for monitoring contractor schedule adherence and shall be responsible for schedule changes due unforeseen issues, work scope expansion or contractor performance. The Project Controls personnel shall review the decommissioning contractors work schedules, budgets, and other relevant documentation. Additional specific responsibilities of the Planning and Scheduling personnel include:

• Review and make recommendations to the Decommissioning Contractor Manager and ACOS/R on contractor work schedules;

• Incorporate the contractor work schedules into a master schedule for the decommissioning project;

• Track budget and schedule progress and prepare reports on progress, variance, and trends;

• Maintain all project documentation; and

• Coordinate vendor work plans and logistics to minimize interference and delays.

9.1.3.3 Licensing The Project Controls and Oversight team shall provide expertise to ensure that all activities are conducted in accordance with federal regulations and the AJBRF License and Technical Specifications. Licensing personnel shall assist the Reactor Facility Manager in ensuring that the AJBRF technical specifications and applicable Federal regulations are adhered to. A primary additional responsibility of licensing personnel is to ensure that all packaging, transportation and disposal of radioactive waste to ensure that these shipments performed by the decommissioning contractor, are conducted in accordance with federal regulations. Additionally, the licensing personnel shall assist the Reactor Facility Manager in the preparation of changes to the AJBRF license if required.

Licensing personnel shall have the authority and responsibility to interrupt or suspend any activity if it is deemed to be in conflict with the AJBRF license or regulations 9.1.3.4 Radiological Engineering Radiological Engineering oversight shall be provided by the Project Controls and Oversight team to assist in ensure that radiation protection controls are 78

appropriately applied to all decommissioning activities and that radiation exposures are maintained ALARA. The Radiological Engineering oversight shall ensure that the decommissioning contractors' QA program related to Radiation Protection is effectively implemented. Additional, specific responsibilities include:

• Evaluating the effectiveness of corrective actions in response to radiological incidents;

• Reviewing and monitoring radioactive waste and material shipments;

• Reviewing and, if necessary, performing engineering and analyses in support of the ALARA program;

• Reviewing survey, sampling and dosimetry reports to identify potential problems or opportunities for further reductions in radiation dose to the workers, hospital staff and public; and

• Reviewing and, if necessary, executing radiation protection procedure changes or prepare new procedures.

Radiological engineering personnel shall have the authority and responsibility to interrupt or suspend any activity if it is deemed to be performing in an unsafe, uncontrolled or improper manner.

9.1.4 Omaha VAMC Industrial Safety Manager The Industrial Safety Manager (ISM) shall be responsible for the development, communication and implementation of a safety culture at the Omaha VAMC and among the decommissioning contractor.

The Safety Manager shall provide oversight of all decommissioning activities and plans to ensure that safe work practices are followed. Specific responsibilities include, at a minimum:

• Review of decontamination and dismantling plans and procedures to ensure proper levels of controls are in place during the evolution of the decommissioning activities, in accordance with the requirements of the Omaha VAMC Safety Program, the AJBRF license and Federal, state and local regulations;

• Evaluation of procedural and RHWP compliance by workers and supervisors;

• Investigating, documenting, and evaluating any injuries that may occur during the decommissioning, and developing and monitoring the implementation of corrective measures;

• Assist in the identification of hazardous materials within the work boundary of the AJBRF and maintain an inventory of all such materials;

• Review all RHWPs prior to issuance; and 79

Inspect and monitor work sites and worker performance to identify any hazards and ensure procedural and safety program compliance.

The Safety Manager has the authority and responsibility to stop or suspend any activity deemed unsafe or lacking in adequate safety controls and measures.

The Safety Manager is an ex-officio member of the RSC and shall report to the Director on all decommissioning matters.

9.1.5 Reactor Facility Manager The Reactor Facility Manager has the responsibility to ensure that the requirements of the AJBRF license and technical specifications are compiled, including performance of necessary surveillance, testing, and inspections. The Reactor Facility Manager will review all decontamination and dismantlement plans to ensure that the reactor facility is maintained in a safe and licensed condition, including ensuring that the requirements of the AJBRF security and emergency plans are complied with and maintained. The Reactor Facility Manager shall provide technical oversight during decontamination and removal of reactor components and systems.

9.1.6 Reactor Safeguards Committee The RSC has broad responsibility to provide independent reviews and audits of decommissioning activities for safety and proper controls. The committee will review decommissioning procedures, decommissioning activities dealing with radioactive material and radiological controls as well as review and approve changes to the decommissioning plan. The RSC shall evaluate all changes to determine if un-reviewed safety questions exist. Additionally, the RSC shall perform quarterly audits on safety controls, radiological controls, environmental controls and ALARA plans (to review the adequacy, effectiveness and compliance with regulatory requirements).

At a minimum, the RSC shall meet on a monthly basis based upon the level of decommissioning activities. The RSC shall report to the Director. The Director shall receive the minutes of each meeting and the results of all audits and reviews within forty-eight (48) hours of their completion. The RSC can require work to be interrupted and/or suspended if audit or review findings indicate unsafe or uncontrolled conditions or operations.

The RSC shall provide the following functions:

Review proposed decommissioning plan or procedure changes, including changes in monitoring or control equipment, systems or testing to determine if there are safety questions as defined in 10 CFR 50.59; 80

• Review all new procedures and major revisions having safety significance;

• Oversee and review any operations that could potentially release radioactivity to the environment;

• Review all reports of violations of technical specifications, license and procedures having personnel and/or public safety significance;

• Review NRC inspection reports and responses/corrective actions as necessary;

• Perform audits of plans and programs (such as ALARA, QA, and industrial safety) to ensure compliance with procedures;

• Review of the dosimetry program and radiation exposures to decommissioning workers; Review and approve planned release of radioactive material to the environment;

• Review any unmonitored release or suspected unmonitored release of radioactive material to the environment; and

• Perform audits and review decommissioning tasks and operations, as the Committee deems necessary.

The AJBRF technical specifications and the RSC charter govern these functions.

The autonomous oversight of the RSC is essential for safe decontamination and decommissioning of the facility and the protection of the health and safety of the public. The RSC shall be composed of a minimum of seven members. Four ex officio members are the ACOS/R (Chairman) for Research, Radiation Safety Officer (RSO), the Reactor Facility Manager, and the Omaha VAMC Safety Manager. Other members shall include VA staff and/or decommissioning contractors with specific expertise in the areas of decommissioning, radiological engineering, environmental controls and/or industrial safety.

A quorum requires the presence of not less than one-half of the membership where the operations/decommissioning staff does not constitute a majority.

Qualified alternates may attend meetings in place of a member. However, no more than two alternates may substitute and only one alternate for an ex officio member will represent a quorum. The qualifications of the membership shall be approved by the existing RSC and forwarded to the ACOS/R and Director for approval.

Replacements, additional members and alternates of the RSC shall have their qualifications approved by the RSC and submitted to the Director for approval prior to committee assignment. The qualifications of the ex officio and other members are detailed in Section 9.3 below.

81

9.1.7 Radiation Safety Officer The AJBRF Radiation Safety Officer (RSO) shall have the responsibility and authority for the daily implementation of policies and practices regarding the safe decontamination, demolition and disposition of radioactive material at the AJBRF.

The RSO has the authority and responsibility to suspend any activity involving radioactive material and/or work performed in radiation areas if the methods and/or procedures used are determined to be unsafe. This would include activities that are not in accordance with ALARA principles, may result in an unmonitored or unplanned release of radiation to the environment and/or are contrary to applicable regulations. Responsibilities include:

• Oversee the maintenance, calibration and use of radiation detection instruments, measurement of radiation levels and contamination levels;

• Maintain and implement the ALARA and personnel dosimetry program;

• Direct and evaluate the results of bioassay measurements;

• Maintain an inventory of the radioactive material possessed within the jurisdiction of the AJBRF license;

• Oversee and approve the shipping and receiving of radioactive materials, such as instruments, sources, or special tools;

• Oversee the environmental monitoring programs;

• Provide radiation safety training for all personnel, contractors and staff working at the facility in accordance with the Radiation Protection (RP)

Program;

• Maintain and implement radiological protection procedures;

• Ensure that the activities involving potential radiological exposure are conducted in compliance with the AJBRF license and Procedures, the decommissioning plan, and federal, state and local regulations;

• Review and approve all RHWPs; Oversee the health physics staff and provide technical expertise in the performance of radiation measurement, surveying of contaminated areas and personnel dosimetry during all decommissioning activities and final release survey; and

• Prepare and maintain the documentation required by 10 CFR 20 and the AJBRF RP Program.

The RSO is an ex-officio member of the RSC and advises the RSC about all matters regarding radiation monitoring and radiation safety during decommissioning activities. The RSO reports directly to the Director, Omaha VAMC.

82

9.1.8 Decommissioning Operations Contractor The Decommissioning Operations Contractor (DOC) shall be responsible for the conduct of the decommissioning in accordance with Federal and State regulations, the AJBRF Technical Specifications, and the contractual requirements of the decommissioning contract. The DOC shall provide the personnel and engage the necessary subcontractors to provide all the services and skills required for the decontamination, dismantlement, disposal and re-construction activities. Additionally, the DOC shall provide the radiological and radioactive waste services to support the decommissioning activities, including handling and shipment of radioactive and hazardous waste as well as performance of the decommissioning final survey.

The DOC is responsible for the development and execution of the detailed work plans, policies and procedures for the decommissioning project with ongoing management of all activities and operations. The DOC is responsible for managing the engineering progress, mobilization, decontamination, demolition and removal of materials (including radioactive and hazardous waste) and the final release survey. The Decommissioning Operations Contractor is also responsible for preparation of the final report in support of the license termination applications to the NRC. Specific responsibilities include, but are not limited to:

• Directing and supervising D&D activities and resolving work problems;

• Coordinate, verify, and validate the proper implementation of work control and oversight programs;

• Develop and administer the formats for regular reporting requirements;

• Manage Decontamination and Dismantlement Services and Radiological and Radioactive Waster Services Contractors;

• Support the analysis and reporting requirements of the RSC; and

• Ensure that the QA program is effectively implemented.

The DOC on-site supervisor shall report directly to the ACOS/R and has the authority and responsibility to stop or suspend work of any personnel performing decommissioning work.

9.1.8.1 Decontamination and Dismantlement Services The decontamination and dismantlement services shall include the execution of the decommissioning plans in accordance with the engineering and radiological control procedures, and RHWPs. Specific services shall include the decontamination, cleanup, dismantlement, removal and preparation for disposal of all materials that comprised the AJBRF. This will require providing the skilled 83

labor, supervision, decontamination technicians, and equipment either directly or via subcontractors.

9.1.8.2 Radiological and Radioactive Waste Services Contractor Radiological and radioactive waster services shall include the radiation protection personnel, instrumentation, training, and controls in support of the AJBRF decommissioning. Specific services shall also include provision of personnel, equipment, and transport services for the removal of radioactive waste from the AJBRF, and the arrangements for disposal at licensed facilities. These support services shall be conducted in accordance with Federal and State regulations, and the AJBRF license and radiation protection program.

The radiological and radioactive waste services shall also include the management and technical personnel, equipment, and procedures for the completion of the AJBRF termination survey. The Final Survey supervisor shall report to the DOC and shall advise on the adequacy of the decontamination effort prior to the performance of the final radiation survey.

9.2 Decommissioning Task Management Decommissioning task management will consist of two components: a project level work plan and the processes for controlling daily work activities. The development of detailed specific plans for the decontamination, demolition, radioactive and chemical/hazardous waste removal and final survey will be the basis for bids supplied by qualified contractors to the VA. These plans provide a detailed work breakdown that will allow the Omaha VAMC management, radiological control personnel, and the QOT to understand the sequence and timing of tasks that require oversight and support.

Daily work control will be via the Omaha VAMC RHWP program. In addition to controlling personnel exposure to radiation and radioactive materials, specific task details and instructions will be identified and attached to the RHWP request for that work. Work techniques will be specified to ensure that exposure rates for all personnel are maintained ALARA, and to ensure that the work techniques are suitable and effective in accomplishing the tasks specified. RHWPs will not replace work procedures, but will act as a supplement and will provide a daily accounting for work hours, equipment usage and exposure.

The Decommissioning Contractor Manager shall prepare the RHWP requests along with a detailed breakdown of planned tasks. These requests will include the areas or components affected, the prerequisites (including support equipment, site conditions, environment, personnel and equipment required),

estimated hours to complete the tasks, and previous task dependencies.

84

Additionally, the appropriate work procedures will be utilized for welding, cutting, rigging, decontamination, and other required tasks. The procedures will also be identified on the RHWP request. The RHWP requests shall be prepared for a period of one week and shall be submitted to the Decommissioning Contractor Manager for approval no later than the Monday prior to the week of anticipated start date.

The approved RHWP request(s) shall be reviewed and approved by the QOT before submittal to the Safety Manager. Upon Safety Manager approval, the RHWP will be sent to the RSO who will prepare the RHWP(s) for the specific tasks. RHWPs shall be prepared, reviewed and approved no later than one day prior to their use. After RSO review and approval, the work crews shall review the RHWP requirements, work plan and procedures in the pre-job briefing. This briefing is required prior to commencement of actual work. The purpose of the pre-job briefing is to review the radiological conditions and work plans/procedures. These briefings shall be conducted on each day or shift of multiple day or multiple shift tasks worked under an existing RHWP.

The Decommissioning Safeguards Committee shall review and approve all RHWPs that have an estimated collective total exposure of 2.5 person-rem or greater.

UAII hands tail-gate" meetings shall be held weekly for all project personnel and shall entail the review of documented deficiencies, close calls, industry events and radiological or safety procedure/requirement changes. Senior management personnel (ACOS/R, Safety Manager, and QOT members) shall chair one tailgate meeting per month each to have an open question and answer session and to emphasize safety and ALARA.

9.3 Decommissioning Management and Oversight Positions Qualifications The minimum qualifications and requirements of the AJBRF decommissioning management and oversight team are provided in Table 9.1, Positions and Qualifications below. Additional knowledge and skills requirements may be required at the discretion of the Omaha VAMC ACOS/R.

85

Table 9.1, Positions and Qualifications Position Minimum Requirements.--.

Quality Oversight

• Bachelor's degree in engineering or science.

Team (QOT)

• Minimum of five years experience in supervising or providing oversight of nuclear reactor activities.

• Minimum of ten years experience in health physics, Quality Assurance, operations, or engineering.

Omaha VAMC

• Ten years experience in safety engineering, safety Safety Manager analysis/investigations, and environmental safety analysis.

• Four-year degree in science, industrial hygiene or health related field or five years additional experience.

• Two years experience in safety oversight or supervision.

• Training in and familiarity with OSHA requirements, HAZMAT, and EPR requirements.

• A combination of partial compliance of the above criteria at the discretion of the Director, Omaha VAMC.

Reactor

• Minimum seven years of professional work Safeguards experience in medicine, science or engineering.

Committee (RSC)

• Bachelor's degree in science or engineering or a minimum of five additional years experience.

• Fundamental knowledge of radiation and radiological controls.

• Combination of backgrounds, providing additional expertise and opinions in areas.

• Knowledge of the AJBRF radiation protection program and license requirements.

• Knowledge of industrial safety requirements applicable at the Omaha VAMC facility 86

Table 9.1, Positions and Qualifications (con't)

Position Minimum Requirements Decommissioning

• Bachelor's degree in engineering.

Operations

• Ten years experience in the nuclear industry in the Contractor (DOC) areas of engineering, construction, health physics, Manager decommissioning and/or operations.

• Five years of project management experience in the nuclear industry.

Project Planning

• Two-year post-secondary education or training in and Controls technology.

• Five years experience in project planning, outage planning or scheduling in the nuclear reactor facility environment.

• Competency in and training on commonly used project-planning software (i.e., Microsoft Project, Primavera, P3, etc.).

• Knowledge of reactor systems and operations.

• Competency in office software such as spreadsheets, word processing and basic database skills.

• A combination of partial compliance to the above criteria at the discretion of the ACOS/R.

Radiological

• Four-year degree in engineering or science.

Engineering

• Industry training or certification in health physics, radiation protection and/or industrial safety.

• Five years experience in operations, radiation protection or engineering in a nuclear reactor facility environment.

• Substantial knowledge of the reactor operations, regulatory requirements and industry practices.

• A combination of partial compliance to the above criteria at the discretion of the ACOS/R.

Licensing

• Four-year degree in engineering or science.

• Five years experience in nuclear power plant operations, licensing, engineering, or QAJQC.

• Substantial knowledge of the reactor operations, regulatory requirements, and industry practices.

A combination of partial compliance to the above criteria at the discretion of the ACOS/R.

87

Table 9.1, Positions and Qualifications (con't)

Position MNinimum'Requiremnents.,"-

Reactor Facility

• Two-year degree in engineering, science, or Manager technology, or four years experience in a technical field.

• Five years experience in the operation and maintenance of a research or commercial nuclear reactor.

Radiation

• Two-year degree in health physics, safety or Protection technology.

Technicians

• Meets the requirements of ANSI N3.1.

• Three years experience in radiation protection at a commercial or research/test reactor or part 30 licensee.

• Medical certification for wearing respiratory protection.

Radiation Safety

• Bachelor's degree in Health Physics or related Officer (RSO) field.

• Seven years experience in providing applied Health Physics in commercial or research reactor environment.

• Three years supervisory experience in health physics.

• A combination of partial compliance to the above criteria at the discretion of the ACOS/R.

9.4 Training Individuals (employees, contractors and visitors) who require access to the AJBRF Controlled Access Areas (CM) will receive training in radiation protection and industrial hazards in accordance with the AJBRF Radiation Protection procedures and the Omaha VAMC Safety Manual.

Health physics, industrial safety and health criteria (as well as other standards that guide the activities of the decommissioning plan described below) are discussed in detail in Section 10 below.

88

Training will consist of formal training sessions conducted upon initial employment or on-site arrival for contractors and informal training sessions conducted during weekly tail-gate sessions for all AJBRF contractors and staff.

9.4.1 Radiation Worker Training Radiation protection training provides the necessary information for workers to implement sound radiation protection practices and is required for personnel working in restricted areas. Radiation protection training will be provided to all personnel who will be performing work in radiological areas or handling radioactive materials. The training will ensure that decommissioning project personnel have sufficient knowledge to perform work activities in accordance with the requirements of 10 CFR.1 9.12 and the AJBRF Radiation Protection program. The principle objective of the training program is to ensure that personnel understand the responsibilities for minimizing exposure to radiation, required techniques for safe handling of radioactive materials, and for minimizing exposure to radiation.

General Radiation Worker Training (RWT) will be required for decommissioning project personnel working in restricted areas and will be commensurate with the duties and responsibilities being performed. Personnel completing RWT are required to achieve a passing grade of 80% on a written examination on the material presented. Completion of this training qualifies an individual for unescorted access to radiologically controlled areas. RWT will include the following:

• Fundamentals of radiation;

• Biological effects of radiation;

• External radiation exposure limits and controls;

• Internal radiation limits and controls;

• Contamination limits and controls;

• ALARA philosophy and program;

• Respiratory Protection requirements, philosophy, and program;

• Management an control of radioactive waste, including waste minimization practices;

• Response to emergencies; and

• Worker rights and responsibilities.

In addition to a presentation of the topics identified above, participants in RWT are required to participate in the following demonstrations:

The proper procedures for donning and removing a complete set of protective clothing (excluding respiratory protection equipment);

89

• The ability to read and interpret self-reading and/or electronic dosimeters;

• The proper procedures for entering and exiting a contaminated area, including use of proper frisking techniques; and

• An understanding of the use of RHWP by working within the requirements of a given RHWP.

The following are examples of the training programs applicable to decommissioning activities:

• Safety controls, including: personnel monitoring, radiation surveillance and monitoring of controlled areas, ventilation, posting of hazardous and/or radiological areas, access controls and health physics administrative controls;

• Proper use of personal monitoring devices; and

• Job-specific training, including mockup simulation or pre-performance briefings to ensure proper equipment usage and achievement of ALARA.

Personnel who have documented equivalent RWT from another site may be exempted from participating in training (except for site-specific training on administrative limits and emergency response) but will be required to pass a written examination and attend demonstration exercises.

Records of training will be maintained in accordance with the requirements of the AJBRF Radiation Protection Program and will include each trainee's name, dates of training, type of training, test results, authorization for protective equipment use and instructor's name(s).

9.4.2 Respiratory Protection Training Personnel whose work assignments require the use of respiratory protection devices will receive respiratory protection training in the devices and techniques that will be required for use. The training program will follow the requirements of 10 CFR 20 and will consist of a lecture session and a simulated work session.

Personnel who have documented equivalent respiratory protection training may be waived from this training.

9.4.3 General Site and Safety Training The Safety Manager, the Quality Oversight Team and the Decommissioning Contractor Manager are responsible for ensuring that the decommissioning activities are conducted in accordance with occupational health and safety requirements for both project personnel and the general public. The primary functional responsibility is to ensure compliance with the OSHA requirements and the Omaha VAMC Safety Manual.

90

All personnel working on the decommissioning project will receive health and safety training. The training will provide personnel with the means to recognize and understand the potential risks involved with personnel health and safety issues associated with the work at the facility. Additionally, the training will help to ensure compliance with the requirements of the NRC (10 CFR), the EPA (40 CFR 24.31 1), and OSHA (29 CFR). Workers and regular visitors will be familiarized with plans, procedures and operation of equipment required for safe performance of their work. In addition, each worker must be familiar with procedures that provide for good quality control. Specific responsibilities include conducting an industrial training program to instruct employees in the following:

• General safe work practices;

• Communication of radiological, chemical and demolition-induced hazards;

• Worker rights and responsibilities;

• Use of personnel protection equipment;

• Hearing conservation training; and

• Fall protection.

Tailgate sessions will include training on procedural changes, deficiencies discovered during audits and surveillances, and applicable industry events or occurrences. Informal training will be documented by retention of the sign-in/attendance sheets and the agenda for the particular tailgate meeting. Also, tailgate meetings shall be held in response to any violation of safety or radiation protection procedure/policy violations, lost-time, injuries, or work procedure non-compliances.

9.5 Contractor Support Contractor support of the decommissioning project will fall into two broad areas:

1) assistance with oversight and control of decommissioning tasks and 2) actual performance of the decommissioning tasks.

Oversight shall be primarily the responsibility of the RSC, RSO, and the Safety Manager. Each of these entities or positions shall possess the authority and the responsibility to stop any activity if it is perceived that the work is not performed in accordance with approved procedures, practices, or required by the AJBRF license.

Contractor assistance in project oversight shall be provided as a primary responsibility of the Project Controls and Oversight team, which may include contract personnel with specialized skills or experience not possessed by Omaha VAMC employees. Contract personnel on the QOT shall be hired independently from any of the organizations performing any of the actual decommissioning tasks.

91

The members of the DOC organization shall provide additional oversight. The Radiological and QA personnel shall provide oversight of work plans and procedures to ensure that work is performed in compliance with AJBRF Radiation Protection procedures, the Omaha VAMC Safety Manual and Industrial Safety procedures, and the AJBRF license requirements. The DOC on-site manager shall have the authority and responsibility to stop any work that is perceived to be in violation of procedures, license requirements or the Omaha VAMC Safety Manual.

Sub-contractors may be hired to perform the engineering, decontamination, demolition, waste removal and final termination survey of the AJBRF. Each sub-contractor shall provide an on-site project manager who will report to the DOC on-site manager and who will have the responsibility for conducting work in compliance with AJBRF Radiation Protection program and Omaha VAMC Safety Manual and Industrial Safety procedures. The sub-contractor project managers shall have the authority to stop their work if it is perceived to be contrary to approved procedures, the AJBRF Radiation Protection Program or Omaha VAMC Safety Manual.

Engineering and initial project planning activities may occur at off-site locations.

Ongoing project planning, oversight and management will be conducted at the AJBRF site during the actual decommissioning work, including mobilization and demobilization. Oversight of off-site engineering shall be conducted by periodic audits and surveillance at the vendor site. During the performance of decommissioning activities that require engineering assistance or involvement, the appropriate contractor engineering personnel shall be at the AJBRF.

As stated in Section 9.4, all personnel who will be authorized to enter AJBRF Controlled Access Areas shall receive radiation worker training, respiratory protection training, general site training and safety training. Contractors will be expected to provide training on work procedures and practices, use of specialized tools or equipment, and the contractors' safety and health policies.

Contractor safety and health requirements shall supplement, not precede, those requirements of the AJBRF and the Omaha VAMC.

The DOC on-site manager shall directly oversee the decommissioning work and will conduct daily communications with all contract personnel and on-site supervisors, will conduct weekly tailgate meetings, as well as periodic independent surveillances. The oversight responsibilities of the RSO, Omaha VAMC Safety Manager, RSC, and QOT will also include contractor performance.

All contractors shall be informed that compliance with the AJBRF Radiation Protection program, the Omaha VAMC Safety Manual and Industrial Safety practices, the AJBRF license requirements and all federal, state and local regulatory requirements is a contractual obligation. Contract personnel who knowingly violate AJBRF/Omaha VAMC policies shall be banned from the site.

Contractors' performance will be monitored and evaluated. Those contractors that fail to provide effective management controls and commitments to 92

radiological and industrial safety requirements, as well as compliance with AJBRF license and regulatory requirements will be terminated and replaced.

93

10 Radiation Safety and Health Program 10.1 Radiation Safety Program 10.1.1 Management Policies Decommissioning activities at the AJBRF involving the use and handling of radioactive materials will be conducted in a controlled manner in order to minimize all exposures to radiation. The Omaha VAMC is committed to a radiation safety program that controls radiation dose (internal and external) to workers and members of the public in a manner that avoids unnecessary and accidental doses, and which maintains effluents and doses to workers below regulatory limits.

The foundation of the radiation safety program is based upon several factors: 1) knowledge of the AJBRF environmental impacts and radiation levels prior to decommissioning; 2) detailed results from the site characterization process; and

3) the Decontamination and Decommissioning Plan in respects to anticipated radiation impacts during the actual dismantling process. Based upon these inputs and the overall commitment to maintaining an effective ALARA program, a project-specific radiation safety program will be utilized and integrated into the overall planning for decommissioning activities. The major elements of the AJBRF radiation safety program are described below.

Responsibilities of management and individual workers in carrying out the policy of ALARA are defined in the AJBRF Radiation Safety Manual and implementation procedures. The Radiation Safety Manual provides requirements and guidance in all areas of radiation safety, including management policy, radiation safety training, dose control, contamination control, surveys instrumentation and incident investigation and analysis. The Radiation Safety Manual incorporates industry experience and good practices, requirements of appropriate rules and standards, and guidance from NRC Regulatory Guides, Information Notices, and Bulletins.

Considerations for maintaining radiation ALARA throughout the decommissioning include: 1) a program to keep as much of the reactor area radiologically clean as practicable; 2) an ALARA program requiring planning and the use of dose reduction techniques for all work involving significant radiation exposure; 3) an ongoing training program to ensure that individuals entering and working in radiologically controlled areas can keep their dose ALARA; 4) periodic procedure and work practice reviews to reduce doses further; and 5) a radiation job history file to trend performance and document/retrieve past experience.

94

10.1.2 Radiation Safety Controls and Monitoring For Workers Workplace Air Sampling Program Respiratory protection measures will be employed to protect workers from a variety of airborne hazards. Radiological airborne hazards consist of radioactive particulate materials. Non-radiological hazards include oxygen-deficient atmospheres, airborne asbestos fibers, particulates, and vapors. This program for respiratory protection is based on requirements in 10 CFR 20.1103 for radiation protection and 29 CFR 1910.134 for non-radiological hazards, and complies with the guidelines in the NRC Regulatory Guide 8.15, "Acceptable Programs for Respiratory Protection."

The Respiratory Protection Program includes the following elements as recommended by NUREG 0041, "Manual of Respiratory Protection Against Airborne Radioactive Material":

• Written operating procedures and policy statement;

• Proper selection of equipment, based on the hazard;

• Proper training and instruction of users;

• Proper fitting, use, cleaning, storage, inspection, quality assurance, and maintenance of equipment;

• Appropriate surveillance of work conditions, degree of employee exposure to stress;

• Regular inspection and evaluation to determine the continued program effectiveness;

• Program responsibility vested in one qualified individual;

• An adequate medical surveillance program for respirator users;

• Use of only Bureau of Mines/National Institute of Occupational Safety and Health (NIOSH) certified equipment; and

• Maintenance of a bioassay program.

Assessing work plans, evaluating conditions in the work area, and reviewing available historical data on the airborne hazards for a particular job shall determine respiratory requirements. Respiratory protection equipment shall be selected using allowed respiratory protection factors to ensure that individual limits on intake or exposure are not exceeded.

95

10.1.3 Breathing Zone Air Sampling Applications Localized air samples including breathing zone air samples shall be collected where it is expected that the concentrations of airborne radioactivity are likely to exceed the criteria of 10 CFR 20.1502(b). A grab sample shall be collected when increases in airborne radioactivity are suspected. All air monitoring equipment will be operationally checked daily prior to the start of work at the AJBRF. Continuous Air Monitors (CAMs) or fixed-position air samplers will be pre-staged in areas where dismantlement/demolition work will occur and there is a potential for airborne radioactivity to occur in general areas. Mobile continuous air samplers will be utilized at sites of any work that has the potential for releasing airborne contaminates, such as cutting, grinding, opening of systems, demolishing structures. Additionally, grab air samples will be collected from low volume samplers and analyzed for contamination periodically during a task.

Table 10.1 lists the types of instruments used for air sampling and monitoring.

Section 10.1.9 details the calibration frequency of air monitoring equipment, which is based on vendor recommendations and equipment usage.

96

Table 10.1, AJBRF Radiation Protection Instrumentation Radiation Calibration

.Use.

InstrmentJDeteor-Detector.Type

'Detected,.

Source Direct alpha Ludlum Model 2350 Th-230 (),

and direct wih4394 8 439 proportional Alpha, Beta, or Tc-99 (Psurveys;gBeta 43-94 or 43-10(126 CM2)

Gamma Cs-137 surveys; Beta dtcor 4-06C-37()scans on solid surfaces.

Ludlum Model 2350/ GgeMlerPpTh20()Direct beta SP-1 13-3m or SP-Geiger-Mlter Alpha or Beta Tc99 surveys.

175-3m Dtco Ludlum Model 2350/

Gas-flow T

Direct beta 43-98, or 43-94 or proportional pipe Alpha or Beta T

(99 surveys.

PSL 3R detectors c-(1)

Direct beta Ludlum Model 2350 Shielded Geiger-Beta Tc-99 (13) surveys; Beta with 44-40 detector MOller (15.5 cm2) scans on solid surfaces.

Exposure rate for Eberline R07 Ion Chamber Gamma Cs-137 (y) components in the reactor pool.

Gamma Ludlum Model 2350 exposure rate with 44-2 or 44-10 Nal (TI) Scintillation Gamma Cs-1 37 y (general area) detector and gamma (surface) scans.

Eberline SAC-4 Smear Scaler Counter or ZnS Scintillate Alpha Th-230 (a) counting.

equivalent Ludlum Model 2350 Gamma with 44-2 or 44-10 Nal (TI) Scintillation Gamma ICs-1 37 (y) exposure rate detector and gamma scans.

97

Table 10.1, AJBRF Radiation Protection Instrumentation (con't)

Radiation

';Calibration Use' ZInstrum entDetector-Detector Type

'Detected Sorce Nuclide Canberra Genie PC identification Gamma HPGe Gamma Mixed gamma quantification Spectrometer of soil and sand samples.

Packard Liquid LSC T

H-3 Tritium smear Scintillation Counter L

tum counting.

Eberline Model 3A General Area Continuous Geiger-MOller Alpha, Beta, or Th-230 (a), Tc-airborne Particulate Air Detector Gamma 99 (p),Cs-137 (y) particulate Monitor or equivalent monitoring.

Eberline AMS-3, -4 Shielded Airborne Portable Continuous Proportional Beta or Gamma Tc-99 (P), Cs-particulate Particular Air Monitor Counter/G Detector 137 (y)monitoring at work sites.

Eberline FHT 1659 L Capture air Mobile Air Monitor NBeaT99samples near work areas.

Air sample locations will be chosen as follows:

Samplers will be located in the air flow near the potential or actual contamination release point; Sampler intakes will be located as near to the workers breathing zones as possible without interfering with the work or the worker; and

• More than one sampling point may be appropriate when there are more than one potential or actual release points.

Alarm set points of the continuous air monitors will be set at ten percent Derived Air Concentration (DAC). If monitors or grab samples show airborne contamination greater than ten percent DAC, the work will be suspended and radiation protection personnel will be notified.

98

10.1.4 Respiratory Protection Program.

Use of Process, Engineering Controls, Procedures and Respirators Respirators are to be used to control personnel exposure to airborne radioactive materials when administrative and engineering controls are not effective and the use of respirators result in TEDE being ALARA. The Omaha VAMC Respiratory Protection Program shall comply with the requirements of 10 CFR 20, Subpart H.

Administrative controls will be used to limit personnel access to or time spent in an airborne area. Engineering controls will be used to limit production of airborne contaminants and to control distribution of airborne radioactive materials. Typical engineered controls that will be utilized at the AJBRF to minimize personnel exposure to airborne radioactive materials are:

• Ventilation devices: In-place or portable HEPA filters or Omaha VAMC ventilation systems, local exhaust by use of vacuums;

• Containment devices: Designed containment barriers, containers, plastic bags, tents, and glove-bags; and

• Source term reduction: Application of fixatives prior to handling, misting of surfaces to minimize dust and re-suspension.

Routine operations are planned activities (generally repetitive and occur with various frequencies). For such operations, potential sources of airborne contamination shall be identified so that respiratory protection may be accomplished by the use of process, decontamination, containment, and ventilation measures and by preplanning of work. The use of respirators as a substitute for engineering controls in routine operations is inappropriate.

Non-routine operations are activities that are either non-repetitive or else occur so infrequently that adequate limitation of exposures by engineering controls is impractical. To the extent that process, containment, and ventilation controls are not reasonably feasible in non-routine operations, the use of respirators to avoid excessive exposure to airborne contamination is appropriate.

Emergencies are unplanned events characterized by risks sufficient to require immediate action to avoid or mitigate an abrupt or rapidly deteriorating situation.

Although emergencies are unplanned, preparations will be made for coping with potential emergencies by providing necessary and sufficient respiratory protection for use in potential emergencies that are likely to entail respiratory hazards. Use of respiratory protection equipment during emergency conditions shall be in accordance with established procedures and shall be utilized so as to not hinder major medical or accident mitigation activities.

Only National Institute for Occupational Safety and Health/Mine Safety and Health Administration (NIOSH/MSHA) certified 'equipment will be used in accordance with the requirements of 10 CFR 20.

99

Respiratory protection shall be required when personnel are exposed to an atmosphere immediately dangerous to life or health. Respirators may also be prescribed when personnel may be exposed to non-radioactive contaminants at, or in excess of, the appropriate Threshold Limit Value (TLV) as established by the American Conference of Government and Industrial Hygienists (ACGIH). As with the radiological use of respiratory protection equipment, primary reliance will be placed upon administrative controls to limit personnel access to hazardous areas and engineering controls to limit production of toxic or nuisance atmospheres or to 'clean up" contaminated atmospheres. When such methods are not practical, respiratory protection equipment will be used as necessary.

Respiratory Protection Selection Considerations Respiratory protection equipment will normally be selected that has a protection factor greater than the anticipated peak airborne concentration expressed as a multiple of total DAC. Respiratory protection equipment may be selected that has a protection factor less than the anticipated peak airborne concentration expressed as a multiple of total DAC provided that use of that equipment is expected to result in a lower TEDE. This evaluation shall be documented in the RHWP package. Protection factors for respiratory protection equipment shall be assigned in accordance with Appendix A of 10 CFR 20.1103 criteria.

The determination of respiratory protection requirements and selection of equipment shall be made by trained and qualified individuals only. Training and qualifications will be documented. The selection of respiratory protection equipment therefore requires knowledge of such factors as:

• The chemical, physical, and toxicological properties of the substance against which protection is required;

• The processes occurring during the work activity and conditions of the work area, as they relate to the dissemination of contaminants;

• The potential existence of conditions immediately dangerous to life or health exist or whether health effects would result only after prolonged or repeated exposures,

• The nature of duties to be performed by the user, particularly as they relate to restriction of movements and worker efficiency;

• An understanding of the principles, design, scope of use, limitations, advantages, and disadvantages of the equipment; and

• The external radiation hazards.

100

Medical and Training Requirements for Use of Respiratory Protection Equipment Personnel who require the use of respiratory protection equipment shall receive a physical examination, and be certified by a physician as qualified to wear respiratory protective equipment prior to wearing any respiratory protection device. Personnel shall be medically evaluated to ensure that they possess the physical and psychological capabilities necessary to perform tasks while wearing a respirator. This medical evaluation shall use 29 CFR 1910.134 and NRC Regulatory Guide 8.15, "Acceptable Programs for Respiratory Protection" as guidance in determining if an individual is medically qualified to wear respiratory protection equipment.

Respirator users shall be trained at least annually (every eleven to thirteen months) in the proper use and maintenance of respiratory protection equipment.

A knowledgeable instructor will perform the training of personnel in the use of respiratory protective equipment. The instructor shall have a thorough knowledge of the application and use of respiratory protective equipment and the hazards associated with radioactive airborne contaminants. Training shall include, but is not limited to, the following:

• Presentation of the AJBRF respiratory protection policy statement;

• Discussion of the airborne contaminants against which the wearer is to be protected;

• Discussion of the construction, operating principles and limitations of the various respirators;

• Explanation of why more positive control measures are not always feasible.

This shall include recognition that every reasonable effort is being made to reduce or eliminate the need for respirators;

• Instructions for assuring that respirators are in proper working condition;

• Instructions in donning and removing the respirator properly;

• Instructions in the proper method for checking to ensure an adequate face to face piece seal;

• Instruction in the proper use, maintenance, and storage of the respirators, including assurance to the user that he or she will be issued a sanitized and properly operating respirator;

• Discussion of the types of cartridges and filters commonly used and the application for each type;

• Instruction in emergency action to be taken in the event of a malfunction, including the appraisal of possible hazard in the event the respirator were to be removed in the hazardous area; and

• Discussion of medical signs and symptoms that may limit or prevent effective use of respirators.

101

Maintenance and Storage of Respiratory Equipment Personnel responsible for the cleaning and maintenance of respiratory protection equipment shall receive documented training necessary for fulfilling their responsibilities.

Respirators shall be cleaned and disinfected after each use and if any respirator shows evidence of excessive wear or has failed inspection it will be replaced.

Respirators shall be inspected after each cleaning. Storage locations shall be clean, dry, and free from chemical or physical agents that may be harmful to the respirator construction materials.

Respirators ready for issue shall have no detectable beta gamma or alpha loose surface contamination. Fixed contamination shall be <100 cpm beta/gamma above background on sealing and interior surfaces, and <500 cpm beta/gamma above background on the exterior surfaces.

Procedures The conduct of the respiratory protection program will be through written and approved procedures. Contractor procedures shall comply with the requirements of the AJBRF radiation protection procedures and management policies Procedures shall be reviewed and approved by the RSC and their use monitored by the QOT and the RSO.

10.1.5 Internal Exposure Determination Internal exposure monitoring will be conducted in accordance with approved procedures. Internal exposure evaluation will be performed on an annual basis, at a minimum, or whenever significant changes in planned work evolutions warrant it. A comprehensive air-sampling program will be conducted at the Omaha VAMC to evaluate worker exposures regardless of whether internal monitoring is specified. The results of this air-sampling program will be utilized to ensure validity of specified internal monitoring requirements for decommissioning personnel. If at any time during the decommissioning hazards are encountered that may not be readily detected by the preceding measures, then special measures/bioassay, as appropriate, will be instituted to ensure the proper surveillance of worker internal exposure.

Monitoring for the intake of radioactive material is required by 10 CFR 20.1502(b) if the intake is likely to exceed 0.1 Annual Limit on Intake (ALI) during the year for an adult worker, or if the committed effective dose equivalent is likely to exceed 0.10 rem (1.0 mSv) for a Declared Pregnant Woman (DPW). Prospective 102

estimates of worker intakes and air concentrations used to establish monitoring requirements will be based on consideration of the following:

• The quantity of material(s) handled;

• The ALI for the nuclides of interest;

• The release fraction for the radioactive material(s) based upon its physical form and use; and

• The type of confinement being used for the material(s) being handled.

10 CFR 20.1204 (a) states "For purposes of assessing dose used to determine compliance with occupational dose equivalent limits, the licensee shall, when required under 10 CFR 20.1502, take suitable and timely measurements of (1) concentrations of radioactive materials in air in work areas; or (2) quantities of radionuclides in the body; or (3) quantities of radionuclides excreted from the body; or (4) combinations of these measurements." The AJBRF radiation protection program shall include air sampling and in vitro bioassay and/or in vivo bioassay monitoring in order to comply with this requirement.

Personnel with the greatest potential for intake of radioactive material will be sampled at a frequency determined by the RSO and based on the pulmonary retention class (days, weeks, years) of the radionuclides of concern to evaluate the effectiveness of the air-monitoring program. If the use of engineering and administrative controls are insufficient and respiratory protection equipment is required to be used for protection against airborne radioactive material, air monitoring and bioassays will be performed to evaluate actual intakes in accordance with the requirements of 10 CFR 20.1703 (a)(3)(ii).

Internal monitoring shall be performed by in vitro and/or in vivo monitoring methods. In vitro monitoring (urinalysis or fecal analysis) shall be used to monitor tritium, carbon-14, uranium and transuranic radionuclide concentrations and their elimination rates from the body. In vivo monitoring (whole body counting or lung counting) or in vitro may be used to monitor personnel for radionuclides that emit gamma rays or x-rays. Internal monitoring shall be performed for all AJBRF employees and contractors unless a prospective evaluation has determined that monitoring is not required.

Baseline in vivo and/or in vitro monitoring shall be performed for all AJBRF employees subject to occupational exposure upon start of the decommissioning phase. Baseline in vivo and/or in vitro bioassays shall be performed for all contractors engaged to work in the CM. In addition, in vivo and/or in vitro bioassays shall be performed at least annually for all personnel issued primary dosimetry, and shall be performed whenever an intake Ž ten DAC hours may have occurred during a seven-consecutive-day period based on air-sampling data, accident conditions, equipment failure, external contamination, or other conditions. Special bioassay samples may be obtained, based on nasal smears, 103

air sampling data, etc., as directed by the RSO. These samples shall be analyzed for those nuclides to which the individual may have been exposed.

In vivo and/or in vitro bioassay shall be performed upon job completion or termination of individuals who have been issued a primary dosimeter.

Declared Pregnant Woman Exposure Policy Based on recommendations of the National Council on Radiation Protection and Measurements (NCRP) and on regulatory requirements, controls shall be established for the protection of the embryo/fetus during a female workers pregnancy. These controls shall ensure compliance with regulatory requirements and protect the rights of the female worker.

Declaration of pregnancy is entirely at the discretion of the woman (medical proof is not required). To declare pregnancy, the woman shall inform the RSO, in writing, of the pregnancy and an estimated date of conception so that the estimated dose to the embryo/fetus prior to declaration can be determined. A DPW shall not be permitted to enter airborne radioactivity areas nor be assigned to tasks that could lead to internal radionuclide intakes.

Medical Radionuclide Intake Occupational exposure does not include exposure due to medical administration of radionuclides. Therefore, individuals shall be required to inform the RSO before receiving medical treatments involving radionuclides. After being informed of a medical intake, documentation shall be obtained, signed by the individual stating the date of treatment, radionuclide used, amount of intake, and medical procedure. The RSO shall perform an assessment to determine what work restrictions may be necessary until the medical radionuclides have cleared to avoid problems with frisking/portal monitors, exposure to co-workers, or exposure to external dosimeter.

Internal Dose Assessments An individual's internal dose will normally be determined using the methodology provided in EPA 520/1-88-020 (as sited in NRC Federal Guidance Report No.

11). This report lists the dose equivalent per unit intake, and the values shall be used directly after converting from sieverts per becquerel to rem per microcurie (Sv/Bq x 3.7 x 106 = rem/uCi).

104

Organ-specific committed dose equivalents shall be calculated when the committed effective dose equivalent exceeds one rem or if an overexposure has occurred. The methodology presented in the NRC Federal Guidance Report No.

11 will be used to calculate organ-specific committed dose equivalent.

Each occupational intake of radioactive material that is confirmed by a positive bioassay result shall be investigated and an estimate of the initial intake shall be performed based on standard retention models. When radionuclides with long effective half-lives (Class Y) are internally deposited, the recording and reporting of the internal dose may be delayed for periods up to seven months, unless otherwise required by 10 CFR 20.2202 or 10 CFR 20.2203, to permit the additional bioassay measurements essential to dose assessment.

Intakes through wounds or skin absorption shall be evaluated and, to the extent practical, accounted for in summation of internal and external doses independent of intakes by ingestion or inhalation. The intake through intact skin is already accounted for in the DAC for hydrogen-3 (tritium) and does not need further evaluation. The RSO shall review and approve, on a case-by-case basis, CM access for workers with a treated open wound.

The results of all intake/dose assessments shall be approved and permanently retained with the individual's dose records.

10.1.6 External Exposure Determination 10 CFR 20 establishes a TEDE limit and a total organ dose equivalent (TODE) limit for occupationally exposed individuals. Monitoring of an individual's external radiation dose is required by the regulations if the external occupational dose is likely to exceed ten percent of any dose limit appropriate for the individual.

External radiation monitoring is also required by 10 CFR 20 for any individual entering a high or very high radiation area.

10 CFR 20.1501 (c) states 'All personnel dosimeter (except for direct and indirect reading pocket ionization chambers and those dosimeter used to measure the dose to the extremities) that require processing to determine the radiation dose and that are used to comply with 10 CFR 20.1201, with other applicable provisions of these regulations or with conditions specified in a license or registration must be processed and evaluated by a dosimetry processor: (1) holding current personnel dosimetry accreditation from the National Voluntary Laboratory Accreditation Program (NVLAP) of the National Institute of Standards and Technology; and (2) approved in this accreditation process for the type of radiation included in the NVLAP program that most closely approximates the type of radiation for which the individual wearing the dosimeter is monitored." Only vendors with accredited dosimetry processing services shall be contracted by the AJBRF.

105

The AJBRF shall provide appropriate monitoring for the accurate determination of occupational radiation exposure to personnel who enter and/or work in the restricted area. Official determination of external exposures shall normally be made using primary dosimeter results except as otherwise noted. Any significant positive results (2 ten mrem) are entered into the individual's official dose history.

External Dose Monitoring External radiation dose monitoring is accomplished through the use of a combination of Optically Stimulated Luminescence Dosimeters (OSLs) and Self-Reading Pocket Dosimeter (SRPDs). The official record of external dose to beta, gamma, and neutron radiation shall be obtained from primary dosimetry (e.g.,

OSLs). Secondary dosimetry (e.g., SRPDs) shall be used to track dose between OSL-processing periods and may also be used as a backup to the OSL. The OSLs (or other primary dosimetry devices) shall be capable of measuring the deep dose equivalent (DDE) at a tissue depth of one centimeter, measuring the lens dose equivalent (LDE) at a tissue depth of 0.3 cm, and measuring the skin dose equivalent (SDE) at a tissue depth of 0.007 cm.

Alarming Dosimeters (ADs) shall be used as secondary dosimetry, as deemed necessary by the RSO, due to the nature of the task or area, i.e., work in high radiation areas, or in high contamination areas that make manipulating a SRPD difficult or inconvenient.

OSLs shall be routinely processed on a monthly basis. Special non-routine processing of OSLs shall be at the discretion of the RSO to ensure personnel dose limits are not exceeded.

SRPDs and ADs shall be processed and certified for accuracy prior to initial use and at an interval not to exceed twelve months.

Exposure Limits To provide assurance that individuals do not exceed the federal limits specified in 10 CFR 20, administrative dose control levels will be established for the AJBRF that are less than those limits. When an individual approaches or has reached his/her administrative dose control level, that individual's access to radiation areas shall be restricted to minimize further occupational dose to ensure that the administrative dose control level and the federal limit are not exceeded.

Administrative control of all personnel authorized entry or work in restricted areas of the reactor facility shall be through RHWPs. Access authorization is valid only after surveys and appropriate evaluations relative to the determination of current radiological hazards including dose rates, hot spots, contamination levels and airborne radioactivity concentrations has been performed of the areas to be entered. For each job function and associated task, RHWPs will be generated 106

and issued specifying all protective requirements and any special instructions associated with the specific work function.

No minors will be permitted to visit the AJBRF.

The dose to the embryo/fetus of DPW shall be limited to 500 mrem during the entire time of pregnancy. Additionally, substantial variations in dose rate to a DPW shall be avoided.

Dose Assessment The RSO shall maintain an accumulated dose for workers as determined by both OSLs and SRPDs. The results of primary and secondary dosimeter shall be compared for each monitoring period. An investigation will be conducted whenever the dose indicated by either the primary dosimeter reading or the secondary (SRPD cumulated) dosimetry readings for the corresponding exposure period is greater then 300 mrem and the percent difference between primary and secondary readings exceeds 25%.

Dose assessments shall be reviewed and approved by the RSO prior to assigning a dose other than that measured by a primary dosimetry device.

10.1.7 Summation of Internal and External Doses Internal and external doses shall be summed whenever positive doses are measured. The dose to the lens of the eye, skin, and extremities are not included in the summation. Compliance with the summation shall be demonstrated by showing that one of the following conditions are met at a minimum:

• The deep dose equivalent divided by five rem plus the sum of the fractions of the inhalation ALI values for each radionuclide does not exceed one; and

• The deep dose equivalent divided by five rem plus the total number of DAC-hours for all radionuclides divided by 2000 does not exceed one.

• The deep dose equivalent divided by five rem plus the sum of the effective dose equivalents to all significantly irradiated organs or tissues divided by five rem does not exceed one.

An organ or tissue is deemed to be significantly irradiated if, for that organ or tissue, the committed effective dose equivalent is greater than ten percent of the maximum committed effective dose equivalent for any organ or tissue. Per 10 CFR 20.1003, Committed effective dose equivalent is the sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the committed dose equivalent to these organs or tissues.

107

10.1.8 Contamination Control Radioactive material and contamination control measures shall be established to prevent the spread of contamination to clean areas, minimize the need for respiratory protection devices, and maintain personnel exposures (internal and external) ALARA. The primary means of preventing the spread of contamination is to contain it at its source and to minimize the number of contaminated areas and the amount of loose surface contamination in those areas. Control of radioactive contamination is accomplished by:

• Identifying and minimizing sources of contamination and radioactive materials;

• Evaluating radioactive contamination survey results to determine the appropriate type and level of personnel protective equipment;

• Establishing limits in operating procedures for radioactive contamination levels;

• Establishing boundaries for contaminated areas;

• Planning and performing work to minimize the spread of contamination to areas and personnel including the use of containments when practical;

• Monitoring personnel, material, and equipment as soon as possible as they leave contaminated areas and CM; and

• Implementing effective "good housekeeping" practices to minimize the size and number of contaminated areas.

Administrative Controls/Procedures Controls shall be applied to any area where removable contamination is in excess of the limits listed in NRC Regulatory Guide 1.86 and any area posted as an airborne radioactivity area. The controls shall include conspicuously identifying and controlling access to contaminated areas.

Access to a contaminated area will require a RHWP and will be controlled by radiation protection personnel at the entry areas. Levels of contamination shall be identified that require assignment of stay times, respiratory protection, alarming dosimetry and supplemental area and/or air monitoring.

Minimization of the potential for spread of contamination will be accomplished by instituting work practices such as:

Making use of tools or equipment which are within a contaminated area rather than introducing additional tools or equipment;

• Bagging, sleeving, covering, or coating tools and equipment prior to bringing items into a contaminated area; 108

• Minimizing materials brought into a contaminated area, such as packing materials, papers, extra or backup tools and equipment;

• Avoiding the use of wooden pallets or other hard-to-decontaminate materials;

• Removing unneeded equipment, tools, materials from contaminated areas as soon as practical; and

• Avoiding the use of a contaminated area for storage of non-radioactive materials.

Surveys Routine contamination surveys shall be conducted at frequencies and locations dependent upon the work activity performed.

At a minimum, contamination surveys of AJBRF shall be performed on a weekly basis to monitor the potential spread of contamination and to identify areas that warrant decontamination. Decontamination will be performed either as "good housekeeping" or due to an unusually high level of beta/gamma contamination or any detectable alpha contamination.

Non-routine contamination surveys are conducted as deemed necessary by the RSO to detect the presence of, or prevent the spread of contamination and, as necessary, to prepare RHWPs and monitor associated work. These surveys shall be used to ensure proper contamination control and respiratory protective measures are being applied. Enhanced personnel contamination surveys will be required (e.g. nasal and face smears, sampling of hair) when the work performed results in unexpectedly high levels of loose or airborne contamination.

Baseline surveys to determine the natural background at the AJBRF were performed during the site characterization survey (Attachment A). Follow-up background surveys, in the areas immediately outside the AJBRF boundary shall be performed monthly. Areas surveyed will include the adjacent rooms, hallways in the Omaha VAMC basement and the areas outside the controlled area at the access to the AJBRF access ramp.

Contamination Action Levels It is the policy of the Omaha VAMC management that detectable contamination on personnel be maintained ALARA. The individual site conditions and isotopes of concern will dictate the monitoring requirements. The contamination limits shall be specified in the AJBRF radiation protection procedures.

Areas shall be identified and controlled as contaminated when removable contamination levels exceed 1000 dpm/100 cm2 of beta/gamma-emitting 109

radionuclides or 20 dpm/100 cm2 of alpha-emitting radionuclides. Equipment, materials, and tools shall be controlled when total contamination exceeds 100 cpm above background using a detector at least as sensitive as a pancake GM detector or removable contamination exceeds 1000 dpm/1 00 cm2 of beta/gamma-emitting radionuclides or 20 dpm/100 cm2 of alpha-emitting radionuclides. Internal surfaces that have been exposed to radioactive contamination shall also meet these limits.

The fixed contamination limit allowing protective clothing to be reused shall be 42,000 dpm/100 cm2 averaged over 300 cm2 and for scrubs shall be 5,000 dpml100 cm2 averaged over 300 cm2. Clothing that exceeds this limit shall be removed from service and discarded as radioactive waste or stored separately for use as an outer layer of protective clothing while working in high contamination areas or mixed hazardous areas requiring multiple layers of protective clothing and shall be discarded after such use.

The decision to decontaminate an area, system or equipment, to require additional protective clothing/equipment (respirators, HEPA filters), or to obtain special tooling will be based on an ALARA analysis that considers exposure reduction, radioactive waste generated, changes in time required for the job, and decontamination of equipment. The RSO shall determine the need for an evaluation of a task either during the review of the RHWP or during the job as determined by job progress, personnel exposure/contamination and/or feedback from the workers.

Source Leak Testing The control of sources shall be in accordance with the requirements of 10 CFR 20, the AJBRF Radiation Safety Manager and implementing radiation protection procedures. Procedural requirements require periodic leak-testing of sources used at the AJBRF.

Routinely, each source considered to be in service will be leak-tested monthly.

Additionally, all sources will be physically inventoried monthly and their physical condition and location will be noted. If a source is found to be damaged, it will be administratively and physically removed from service, leak-tested, and the RSO notified.

Sources that are administratively removed from service shall be tested upon removal from storage for disposal or for use.

Smears shall be performed on each source, unless the source is electroplated or has a mylar covering. Additionally, smears shall be performed on the source container. Smears will be counted for beta-gamma and/or alpha contamination using appropriate instrumentation and adequate MDA values.

110

Sources indicating smear results greater than procedure specifications shall be removed from service and contained in plastic or other material to prevent the potential spread of contamination. The RSO shall be notified and the source decontaminated, if possible, and the source shall be returned to the manufacturer or vendor for repair or disposed of as radioactive waste.

10.1.9 Radiation Monitoring Instrumentation Program Various types of radiological and hazardous measurement instrumentation will be used at the AJBRF for radiation protection and hazardous monitoring purposes.

The AJBRF instrumentation program will include procedures for inventory, issuance and control, calibration, operation, response testing, maintenance, repair, and quality control of radiation protection and hazardous instrumentation and equipment. The procedures will ensure compliance with applicable federal regulations and Omaha VAMC policies.

Instrumentation and Equipment The decommissioning vendor shall provide radiation-monitoring equipment utilized during the AJBRF decommissioning. However, all contractor supplied equipment shall be controlled, maintained, tested, calibrated, repaired, and operated in accordance with approved AJBRF radiation protection procedures and the requirements of this plan. The QOT shall audit vendor performance and adherence to approved procedures. Daily oversight and operational/response checks of equipment shall be provided by the RSO.

Instrument Inventory and Control Issue, control and accountability of radiation protection and hazardous instrumentation will be performed in accordance with formally established implementation procedures and will be consistent with regulatory requirements.

A sufficient inventory and variety of operable and calibrated portable, semi-portable and fixed radiological protection and hazardous instrumentation will be maintained. A listing of the types, quantities, and capabilities of the planned instrumentation to be utilized in support of the decontamination and decommission effort is depicted in Table 10.1, above.

Instrumentation and equipment will be identified and tracked by means of a serialized inventory system. At a minimum, this inventory will provide for a distinct identifying number, the most recent calibration date, the next calibration due date, and the status of the instrument (out for repair, ready issue, etc.).

111

Assigned radiation protection personnel will document issuance and return of radiation protection and hazardous instrumentation.

Calibration Calibration of radiation monitoring, counting and air sampling instruments will be performed in accordance with ANSI N323-1978, 'Radiation Protection Instrumentation Test and Calibration," and will be performed in accordance with specific written procedures that reflect current regulatory requirements and industry standards.

All instruments will be calibrated prior to first use, following any repair, maintenance, or modification that could affect calibration, after failure of a response test requiring adjustments. The calibration frequency for portable radiation monitoring instruments and portable air sampling equipment will be at least every six months. Semi-portable (e.g., continuous air monitors, personnel contamination monitors) and fixed (e.g., count room/laboratory instrumentation, portal monitors) instrumentation will be calibrated at least annually.

Approved instrument vendors, at their respective facilities, will provide calibration of all radiation protection and hazardous monitoring instrumentation. Calibration certifications supplied by vendors and traceability of calibration sources will be reviewed for completeness and accuracy prior to accepting instrumentation for use.

Instrument-specific calibration procedures developed and implemented by vendors will include, as appropriate:

• Instrument specifications and sensitivity for each type of radiation to be measured;

• Operating settings/parameter;

• Determination of Minimum Detectable Concentration (MDC)/Minimum Detectable Activity (MDA);

• Environmental limitations;

• Calibration frequencies and tolerances; and

• Data forms, regulatory and instructional references, and applicable drawings and schematics.

Radiation protection instrumentation and laboratory analysis equipment will be calibrated using National Institute of Standards and Technology traceable sources, or equivalent. The sources will be of the type, energy and geometry representative of the radiation to be measured. Only personnel trained and 112

qualified in the use of applicable procedures and test equipment will perform calibrations.

Maintenance and Repair Approved contractors using qualified personnel will perform maintenance and repair of radiation protection instrumentation. All maintenance and repair will be documented. Due to the relative short time span of the actual decommissioning work, contractor instruments requiring repair shall typically be replaced rather than repaired and returned to use at the AJBRF.

Operation and Use Operation of radiation and hazardous monitoring, counting and air sampling instruments will only be performed by personnel qualified in the use of the instrument. Additionally, operation will be performed in accordance with the operational procedure for each type of instrument in use. Operating procedures will include, as appropriate:

• User responsibilities;

• Instrument and detector description;

• Precautions and limitations;

• Pre-operational and operational instructions;

• Identification of proper check sources and associated jigs;

• Performance of response test and/or operational checks; and

• Methods for documenting response checks.

Response testing of portable radiation monitoring instruments will be performed and documented daily or prior to use on each scale that will be used. Any scale not checked will be clearly labeled.

Quality Control and Quality Assurance (QC/QA)

A QA Program for counting instruments will be established and maintained to ensure reliability of counting results and sensitivities. QC for counting instruments will be formally documented and performed using procedures developed consistent with the guidance set forth in ANSI N323-1978, 'Radiation Protection Instrumentation Test and Calibration" and regulatory requirements.

113

QC of portable survey instruments will be performed by the use of pre-operational and response checks in accordance with specific procedures.

The QOT and/or the RSC will perform QA/QC review and evaluation of the instrumentation program at least quarterly as part of radiation protection audits.

10.2 Nuclear Criticality Safety All fissile material (i.e., fuel and fission chambers) have been removed from the AJBRF and transferred to the USGS GSTR facility. There is no fissile material at the AJBRF and thus no nuclear criticality issues of concern.

10.3 Health Physics Audits, Inspections, and Record Keeping 10.3.1 General Records related to health and safety of radiation workers and individual members of the public shall be maintained as required by procedures and reactor license technical specifications. Record-keeping is important in demonstrating regulatory compliance issues as well as providing means to resolve potential liability issues that may arise. All records shall be prepared and maintained in accordance with the AJBRF radiation safety manual and implementation procedures.

10.3.2 Records Records, such as the Radiation Safety Manual, Radiation Protection Procedures, the AJBRF QA Manual, and other records (including as surveys, training records, assessments, and Decommissioning Safeguards Committee meeting minutes) will be generated and maintained to document the implementation of the AJBRF radiation protection program. Records of the Radiation Protection Program will be retained in accordance with established procedures and reactor license technical specification requirements.

The QOT and the RSC shall periodically audit the radiation protection program records such that an assessment is performed at least quarterly. Records of internal audits (by Omaha VAMC committees, DVA contractors or staff) shall be maintained for five years after license termination. Records of NRC audits and inspections, as well as responses to NRC requests for information, inspections and notices of violation shall be maintained for five years after license termination. Other records such as instrument logs, instrument repairs, procedure changes, and/or notes by the RSO/Radiation Protection Technicians shall be maintained as deemed necessary by the RSO and/or Assistant Chief of Staff for Research.

114

Specific Radiation Protection records that will be generated and maintained for at least five years after license termination are:

• Results of radiation, contamination and airborne surveys;

• Shipping papers for radioactive waste and materials;

• Results of bioassay analyses;

• Calibration records for survey and laboratory instruments;

• Personnel monitoring and exposure records;

• Radiation/Hazardous Work Permits;

• ALARA reviews;

• CAA entry, survey, and turnover logs;

• Results of laboratory analyses of material and samples;

• Source inventories and source leak tests; and

• Training records, including fit tests and medical clearances for respirator use.

Radiation Protection records shall be controlled in accordance with regulatory requirements, the requirements of the AJBRF QA Program Plan and the Radiation Safety Manual and its implementation procedures.

10.3.3 Audits and Assessments The QOT and RSC will conduct assessments periodically so that some portion of the AJBRF radiation protection program shall be audited quarterly. The areas to be audited are:

• Radiation Protection Program Administration;

• Personnel Monitoring Program;

• ALARA Program;

• Radiation/Hazardous Work Permit Program;

• Radiation Protection Survey Program;

• Respiratory Protection Program;

• Radioactive Materials Control Program;

• Unconditional Release Program;

• Contamination Control Program;

• Radiation Protection Instrumentation Program;

• Environmental Monitoring Program; and 115

0 Radioactive Waste Management.

Oversight and surveillance of adherence to the radiation protection daily activities of the radiation protection personnel and workers will be conducted at least weekly during the decommissioning work. Audits and assessments records will include the name of the auditor(s), date of audit, areas audited, findings and corrective actions prescribed, and post corrective action implementation follow up. Records of these audits will be maintained for five years after license termination.

10.3.4 License and NRC Violations Findings during audits by the RSC, QOT, or AJBRF management of violations of the AJBRF license or NRC regulations shall be reported to the NRC in accordance with 10 CFR 20 and the AJBRF license requirements. Reports of violations provided by AJBRF employees or contractors to the AJBRF management shall be investigated by the RSO, Industrial Safety Manager (ISM),

or QOT. Results of the investigations by the RSO/ISM shall be documented as appropriate and presented at the next scheduled RSC meeting. The RSC will be convened earlier than their next routine meeting to review and develop/approve corrective actions at the discretion of the ACOS/R. Documentation of such violations and subsequent investigations and corrective actions will be prepared and maintained.

10.4 General Industrial Safety Program The Omaha VAMC Safety Manager is responsible for ensuring that the decommissioning process meets the Omaha VAMC occupational health and safety requirements applicable to project personnel and the general public. The primary functional of the Safety Manager is to ensure that all applicable OSHA requirements are implemented and that full compliance is achieved.

Responsibilities of the Safety Manager include, but are not limited to, the following:

• Conducting applicable industrial training of personnel in general safe work practices commensurate with the activities to be undertaken during the decommissioning process;

• Reviewing project decontamination and decommissioning procedures to ensure that appropriate industrial hygiene needs and requirements are addressed and adequately incorporated;

• Performing periodic inspections of operations areas and work activities to identify and correct unsafe conditions and work practices;

• Providing industrial hygiene services as required; and 116

• Advising the appropriate Omaha VAMC management on industrial safety matters and on the results of periodic safety inspections.

117

11 Environmental Monitoring and Control Program Decommissioning activities at the AJBRF will be conducted in a controlled manner to minimize both public and occupational radiation exposures. A site-specific environmental monitoring program will serve to protect the workers and public from exposure to radioactive materials. The program is based upon several factors: knowledge of the AJBRF environmental impacts and radioactive levels prior to decommissioning via the site characterization and pre-decommissioning surveys, and the environmental impacts anticipated from the decontamination and dismantling activities detailed in the decommissioning plan.

Based upon these inputs and the overall commitment to maintaining exposure ALARA, a site-specific Environmental Plan will be created and in implemented during the decommissioning. This section describes elements of that plan. The overall objectives of the environmental monitoring and control program include the following (ANSI/Health Physics Society N13.1-1999):

• Meeting regulatory requirements;

• Assessing the need for a permanent sampling or monitoring program;

• Assisting in evaluating claims of radiation injury by workers or others;

• Measuring the release of radioactive materials to the environment through source sampling;

• Helping to ensure that people in the surrounding environment are not exposed to the levels of airborne materials exceeding established limits; and

• Helping to assess the possible consequences of non-routine incidents and guiding the selection of appropriate corrective actions that may include the integration of radioactive contamination released to the environment over various time periods.

The site characterization survey indicated that most structures and surface areas within the AJBRF are not radioactively contaminated. Also, the characterization survey revealed that the areas in the outside environment of the AJBRF are not radioactive above natural background levels. The survey results indicated that the reactor components and laboratory drain lines have levels of radioactivity that will need to be decontaminated or removed as radioactive waste. The removal of these components and materials will be carried out in such a manner that the spread of contamination will be contained and minimized and thereby reduce the chances of elevated releases to the environment.

118

11.1 Environmental As Low As Reasonably Achievable (ALARA) Radiation Exposures Evaluation Program 11.1.1 Management ALARA Commitment The Omaha VAMC management is committed to maintaining occupational and public exposures, as well as effluent discharges ALARA. This will be accomplished through the implementation of a sound radiation protection and environmental monitoring programs including the use of engineering controls and ALARA principles commensurate with the radiation and contamination levels associated with activities performed during the decontamination and decommissioning of the AJBRF.

11.1.2 ALARA Policy The AJBRF decontamination and decommissioning activities will be conducted in such a manner as to achieve ALARA exposures to workers and the public. The basic philosophy of radiation protection is to achieve radiation exposures and effluent releases below applicable regulatory limits. Benefits and risks are considered in attempting to maintain low exposures. The AJBRF Radiation Protection Program endorses the ALARA philosophy through the use of procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are ALARA. The Omaha VAMC management is committed to providing all necessary resources, in the form of personnel, training, engineering controls, preparation and planning, design, equipment, monitoring, devices and controls to achieve ALARA doses within the AJBRF and the surrounding environment.

All employees are expected to be knowledgeable of work activities and to abide by all ALARA requirements documented in work instructions. In addition, each employee is responsible for minimizing dose to himself, workers and members of the public.

Relative to maintaining effluent releases to the environment ALARA, the Omaha VAMC shall maintain a zero release policy for liquids and particulate airborne effluents associated with the AJBRF. Specifically, Omaha VAMC management's position is that no hazardous or radioactive liquids will be further discharged by way of sinks and drains or other means of discharge to the environment during the decommissioning. This philosophy is undertaken to ensure decommissioning efforts and exposures to personnel and members of the public will be minimized.

119

11.1.3 ALARA Goals Investigation Levels: In order to ensure that AJBRF ALARA goals and objectives are met, very conservative investigation levels will be established prior to the decontamination and decommissioning effort and applied via written procedures. As specified in Sub-section 11.3 below, action levels for effluent releases and direct radiation values in uncontrolled areas outside the AJBRF are established at a very low level compared to regulatory limits. Routinely radiation surveys will be conducted in the outside areas adjacent to the AJBRF to ensure that equipment or radioactive materials being stored or staged for shipment do not create unwarranted radiation levels. Air sample from AJBRF emissions will be reviewed routinely to anticipate appreciable increases in effluent discharges and to allow for timely corrective actions.

Oversight: The RSC and QOT will serve as the oversight bodies for all decommissioning activities. This includes auditing and overseeing the radiation protection and the environmental monitoring program. This RSC will meet, at least monthly and as needed, to oversee the implementation of the environmental monitoring program. Members of the QOT will monitor work and radiation protection practices and performance daily. Established radiation protection procedures and reactor technical specifications state when the RSC is to be notified when a radiation protection exposure and/or an effluent administrative limit is exceeded. The RSC shall also approve ALARA goals for the radiation protection, as well as, ALARA goals specific to the decommissioning effort.

The RSC Chair shall report on decommissioning issues that require additional management attention to the Director of the Omaha VAMC.

11.2 Effluent Monitoring Program 11.2.1 Policy, Method, Frequency and Procedures Monitoring will consist of routinely measuring the quantity of direct radiation and radioactive material outside of the AJBRF. The monitoring results will establish the radiological conditions as a function of time, provide for a permanent record of these conditions, and permit evaluation of radiological trends over time. The environmental monitoring program will be instituted prior to commencement of decommissioning activities. The environmental sampling and measurement results obtained during the site characterization survey will serve as the baseline data benchmark for comparison with in-progress and final survey results following completion of the decommissioning efforts.

Direct radiation levels will be monitored with OSLs continuously positioned at pre-determined sample locations in the environment outside the AJBRF. Dose 120

rate in millirem/hr will be computed from the time of installation to time of removal. The OSLs will be processed monthly.

A continuous air particulate fixed-filter monitor with audible and visual alarms will be utilized to monitor the exhaust system from the AJBRF.

11.2.2 Airborne Radioactive Effluent Monitoring The primary method of removing exhaust air from the areas of the AJBRF is through a common currently installed ventilation system that runs through the Omaha VAMC facility and exhausts to the building's 12th floor roof. During reactor operations the primary concern for effluent control purposes was related to short-lived gaseous emitters. However, during the decontamination and decommissioning process the main airborne effluent constituent will consist of particulates. A HEPA filtration system, with the exhaust capacity required by AJBRF Technical Specifications, will be installed and operate in place of the building ventilation system. As depicted on the Figure 2.1, the AJBRF main air effluent discharge will be located on the east outer side of the AJBRF.

Monitoring of this effluent discharge point will be accomplished by installing a portable radiation air monitor at this location for purposes of collection of routine and special air samples from the ventilation exhaust point. The instrumentation to be used for this monitoring will have the capabilities to monitor real time radioactive releases as well as collect samples for laboratory analyses.

Several hoods located in the AJBRF that were previously used in analyzing samples removed from the reactor will be disabled as the site characterization survey results revealed that the ventilation supporting the operation of these hoods were basically free of radioactive contamination. Continued operation of these hoods would potentially allow for contamination and would have to be decommissioned as part of the overall process. The discharge points for these hoods are located on the roof of the hospital complex and will not be further utilized.

Additionally, as an added conservative measure, local engineering controls, such as containments and HEPA filters, will be temporarily installed to minimize the concentrations of airborne radioactivity to general areas within the AJBRF due to decontamination and dismantlement activities. These engineering controls, supported by CAMs, will be utilized commensurate with the radiological hazard anticipated as a result of specialized decommissioning activities.

11.2.3 Liquid Effluent Monitoring As stated above, Omaha VAMC management has mandated a zero release policy for radioactive liquid discharges. Accordingly, there is not expected to be a 121

need for monitoring liquid discharges since there will be no discharge points established or authorized.

The wells established outside and close by the AJBRF during the course of the site characterization will be monitored on a routine basis. Environmental monitoring of these areas will consist of routinely collecting water samples and comparing results to those taken during the site characterization. The site characterization results for these wells, which indicated non-detectable activity, will be used as a basis for demonstrating that radioactive materials from decommissioning activities are being properly contained and monitored.

11.2.4 Unmonitored Effluents If special conditions or circumstances during the decommissioning require release of liquids or airborne effluents to the environment, such releases shall require review and approval by the RSC and authorization from the Director of the Omaha VAMC.

During planning and preparations for the decommissioning work, all drain lines and systems connected to the AJBRF will be evaluated for unmonitored pathways. This includes those drain/sewer lines connecting to other systems within the Omaha VAMC and those drains or lines that discharge directly to public common systems. Air monitoring and/or sampling systems will be installed to sample airborne effluents from the facility ventilation discharge system to ensure airborne contamination does not enter the sealed building ventilation system and exhaust to the environment.

11.3 Effluent Control Program 11.3.1 Administrative Controls As stated above, the objective of the AJBRF management is to maintain a zero release site for both liquid and air effluents. Storage of solid radioactive waste packages will be maintained at a minimum to prevent the potential inadvertent release of radioactivity to the environment. Air and water action levels will be established at an administrative control level of ten percent of the applicable Table 3 criteria contained in Appendix B of 10 CFR 20. Soil and sediment action levels will be established at five picoCi/g for radionuclides other than natural occurring. Ambient gamma levels will be established at 25 millrem per/quarter and one millrem in one hour for individuals in unrestricted areas.

11.3.2 Corrective Actions Administrative controls will require immediate notification to the Radiation Safety Officer for any samples or radiation exposure levels that exceed action levels in 122

sub-section 11.3.1, above. The Radiation Safety Officer will initiate or perform an investigation and response consisting of one or more of the following actions:

• Verify applicable laboratory data and supporting calculations;

• Analyze and review probable causes;

• Determine need for re-analysis or additional analyses on original sample;

• Evaluate the need for re-sampling;

• Evaluate the need for sampling of other pathways;

• Determine the need for notifications and reporting as required by internal procedures, technical specifications and federal requirements;

• Document all actions, analyses and evaluations in applicable logs or files;

• Perform applicable exposure assessments; and

• Develop corrective/preventive action recommendations to prevent future releases.

11.3.3 Leak Detection Systems This area of the environmental program at the AJBRF is considered non-applicable, due to the fact that ponds, lagoons, and tanks do not exist in areas within the AJBRF or in the outside environment. Specifically, there are no closed or open liquid pathways, with the exception of the reactor pool and assorted drain lines within the facility. During the site characterization, several wells were established for purposes of monitoring for radioactivity before, during, and following the reactor decontamination and decommissioning activities. With regard to drain lines located within the facility administrative controls will be established that prohibit the use of the drains, (i.e., discharging of liquids into the drains) and the actual capping of the drains during decontamination/removal of the drains themselves.

11.3.4 Disposal to Sewage Systems Administrative and physical controls will be established and implemented that do not allow for the use of discharges through the sewage systems at the AJBRF.

11.3.5 Quality Assurance Administrative controls will be established to ensure that samples being collected from the discharge areas are representative of the material sampled. Replicate samples will be taken periodically to determine the reproducibility of sampling.

123

Provisions will be established that ensure sample integrity will be maintained from the time of collection to the time of analysis.

As necessary, the utilization of quality control sample analysis will be instituted as a means to determine the precision and accuracy of the monitoring process.

When deemed necessary, the use of intra-laboratory and inter-laboratory sample analyses will be considered to verify the quality of analysis methods.

Audits of the environmental program will be carried out by the QOT on an established schedule associated with the type of activities occurring during the decommissioning project.

124

12 Waste Management Program 12.1 General Waste Management Program 12.1.1 Waste Management Statement The Omaha VAMC is committed to the safe decontamination and decommissioning of the AJBRF. The primary objective of the Omaha VAMC's waste management program is to protect workers, visitors and the environment from potential effects of various radiological and hazardous waste steams. The Omaha VAMC is committed to strict compliance with all federal and state radioactive and hazardous waste handling and disposal requirements.

12.1.2 Waste Management Philosophy A waste management program addressing radioactive, hazardous, and mixed-waste streams will be formally established and implemented during the entire decontamination and decommissioning process.

Programs for the management of waste and pollution prevention will be defined to meet the requirements of the AJBRF license technical specifications applicable Federal and State regulations and will be implemented by written administrative and technical implementation procedures.

12.2 Radioactive Waste Management The AJBRF decontamination and decommissioning processes will require the handling and processing of volumes of radioactive waste typical of a research reactor of this type, to reduce residual radioactivity to a level permitting release of the facility/site for unrestricted use. Equipment and materials that cannot be decontaminated to the level below the radioactive release criteria will be processed as radioactive waste.

AJBRF radiation protection personnel will ensure that appropriate processing, packaging and monitoring of solid and liquid wastes generated during the decommissioning process are performed in accordance with formally approved administrative and technical implementation procedures. These programs and procedures will be maintained and controlled in compliance with the AJBRF technical specification requirements and this plan.

125

12.2.1 Solid Radioactive Waste Management Solid radioactive waste handling at the AJBRF will be divided into three phases:

packaging, on-site material staging, and shipment. Each of these phases will be conducted in strict compliance with the AJBRF technical specifications requirements, Waste Disposition Plan and applicable federal, state, and disposal site requirements. In addition, all waste processing and disposition activities will be completed in accordance with the requirements of the Quality Assurance Program (Section 13).

Solid radioactive waste generated during the decommissioning of the AJBRF will be primarily comprised of low-level radioactive waste. The solid waste will primarily consist of activated aluminum/graphite (from the reactor internals),

concrete and structural materials from walls and, exhaust hoods, steel and concrete/epoxy/gunite from the reactor tank and steel/rust from laboratory drains, and a small volume of mixed consisting of contaminated lead-based paint and contaminated asbestos containing floor tiles. Table 12.1, AJBRF Waste Disposition, lists the various components, activity and volumes expected to be generated as a result of the decommissioning and decontamination process.

126

Table 12.1, AJBRF Waste Disposition I

° 4

' S.,..-..,-..,.,.

Processing8 Disposal Envirocaret, Ldfill

-Volume Burial/Mateaia/Sur-aceWaste Estirrmate Barwell Waste Waste

Area Description.'

(iM) ntaminated SiteCode

Tpe Deon.Technique

-m3)

Estimate(i

3) Essmmate(in3)

Structures A0002 - SW2 -

Tiled Reactor Console Concrete/Glass Area 01F01 Floor 0.37 10.0%

E, L Covers Rad. Trash Covers 0.37 A0002 - SW2 -

Reactor Console Reactor Floor Area 01T01/02 renches 0.05 20.0%

E, L Bare Concrete Scabble (1/2 cm) 0.05 A0002 - SW2 -

eactor Console East Cooling ea 01S04 Pit N/A 10.0%

E Concrete Scabble (1/2 cm) 0.01 A0003 - Area 2 SW2C) Lab eas with Hoods Dispose of entire 1S01 Hood EFN1 3.00 100.0%

E Steel & Plastic ood 3.00 A0004 - Area I SW2F) Source oom 01F01 Floor 50.0%

E Concrete Scabble 0.5 cm 0.30 127

Table 12.1, AJBRF Waste Disposition (con't)

Processing&

Disposal Envi r re Landfill Horizontal Storage Vaults A0004 - Area 1 (Length (SW2F) Source provided is for Wash & Remove, if Room 01P01I 6 Vaults) 0.82 3.0%

E Metal & Lead necessary 0.07 Vertical Storage Vaults A0004 - Area I (Length (SW2F) Source provided is for Wash & Remove, if Room 01P02 en vaults) 0.35 10.0%

E Metal & Lead necessary 0.11 A0004 - Area 2 SW2E) 01S01 Hood EF10 10.0%

E, L Steel & Plastic Dispose of Filters 0.40 4.00 Systems Current Drain Remove entire Lines I/S RCA 0.64 100.0%

E Steel/Plastic drain system 1.28 Pneumatic ransfer System ncluding piping Cut Into 1.2 m rom 30.5 cm ections and sent to above core 0.19 100.0%

E Steel E-Care 0.39 128

Table 12.1, AJBRF Waste Disposition (con't)

Processing &.

DIs -po"s'al E'-_rocre Land"

Volume, Burial': Mate'r aSurfce Waste Estimate BamwellWaste Waste Aa

,Desc'iption (i 3) nCtamln'ated t Cod T

. Technique 3

Estmate (m.

3) timate (m

XntarnnatedHe (o

Ty)

Cut into 1.2 m sections and sent to Reactor E-Care bagged as Coolant System 0.75 100.0%

E Steel asbestos 2.98 Remove resin and section metal to

<25.4 cm in one dimension for Ion Exchanger 1.13 100.0%

E Steel Envirocare disposal 1.69 Five micron Metal & filter ilters 0.02 100.0%

E edia Drain and trash 0.07 Reactor Components Water from Tank N/A 0.0%

S Water Process & Stabilize 2.0 Cut into 1.2 m sections and sent to Central Thimble 0.07 E

Aluminum Bamwell Gunite &

Concrete IS and O/S Tank Concrete/Gunite/ Remove and and Rebar 12.20 100.0%

E Epoxy ispose 18.30 129

Table 12.1, AJBRF Waste Disposition (con't) bisjoa Envirocare Lnfl Volume Burial Murfac a Estimate bamell Waste W

AreaDescsiption

.(i 3).

ntainatd Site d

T e

econ.

nique -

(in3)

Estimate(n

3) Estimate(

3)

Remove and Tank 0.27 100.0%

E Steel dispose Remove and send Control Rods 0.10 100.0%

B Boron Carbide to Bamwell 0.10 Graphite/

Cut Out and Reflector 1.2 100.0%

B Aluminum Dispose 1.17 Put In Cask and Dispose of at Core 0.70 100.0%

B Aluminum amwell 0.70 Bridge 0.42 20.0%

E Metal Wash 0.08 130

12.2.2 Solid Radioactive Waste Processing Radioactive waste processing at the AJBRF will be performed in controlled areas that both minimize the radiation exposure to personnel and the movement of the contaminated equipment and materials. These areas will be controlled and monitored to minimize worker exposure and the spread of contamination to the extent practicable.

Radioactive waste packaging operations will be performed following written procedures that ensure:

• Work is performed under an approved RHWP;

• Specific packaging requirements are identified and performed;

• Quality assurance requirements for packaging operations are followed;

• Appropriate monitoring of dose rates and contamination levels are performed and recorded for each package, and

• Each package is appropriately marked, labeled and inventoried.

Waste packages and packaging will meet the applicable requirements of 49 CFR, 10 CFR 20 and 10 CFR 71 and the disposal facility's criteria for transportation and disposal for each decommissioning waste stream.

Implementation procedures and the Waste Disposition.Plan will provide instructions for determining the 10 CFR 61 waste classification of radioactive waste. Instructions for determining the radionuclide content of a container through a combination of direct measurements, radiation shielding calculations, and the use of appropriate scaling factors will be provided by approved procedures.

It is anticipated that the solid radioactive waste will be sent to an established a radioactive waster processing facility for volume reduction, sorting and decontamination.

12.2.3 Solid Radioactive Waste Storage Awaiting Shipment Solid radioactive waste awaiting shipment will be stored in controlled areas that are posted and away from personnel traffic or work areas. Storage will be in accordance with procedures that will address the radioactive material storage requirements of 10 CFR 20. Periodic inspections of these designated areas and those containers stored within the controlled area boundaries will be performed to ensure that package integrity is maintained.

131

Large packages awaiting shipment may be stored in designated areas outside the actual reactor facility prior to shipment. Appropriate security controls and radiological controls will be implemented, as necessary, to prevent unauthorized access to the area and the radioactive materials stored inside. Additionally, precautionary measures will be implemented to ensure that the components are adequately protected from on-site hazards (e.g., extreme weather conditions and heavy load movement).

12.2.4 Solid Radioactive Waste Shipment Solid radioactive waste will be shipped in accordance with formally established implementation procedures. These procedures will ensure compliance with applicable federal, state, and licensed waste processor/ disposal site requirements. Prior to each shipment of a radioactive material package the quality control requirements of 49 CFR 173.475 will be applied.

It is anticipated that the radioactive material shipments will be completed over public highways. The routing of shipments may vary with weather and highway conditions. Additionally, local and state restrictions pertaining to radioactive material transport may affect some route selections. The carriers contracted by the Omaha VAMC will be responsible for selecting the appropriate route, which must conform to applicable federal, state, and local shipping requirements and be in accordance with Department of Transportation and NRC regulations.

12.2.5 Estimated Volumes of Solid Radioactive Waste Based upon the results of the site characterization survey, which included the development of a Radioactive Waste Disposition Plan, preliminary radioactive waste estimates have been determined. This waste determination includes those structures, systems, and components removed from the AJBRF. This detailed analysis estimated that total of approximately 40 m3 of waste together with no more than six m3 of equipment and material waste generated during the decommissioning will have to be processed as radioactive waste. The determination of the amount of radioactive waste due to the contaminated systems and structures within the AJBRF are provided in Table 12.1.

12.2.6 Liquid Radioactive Waste Processing Radioactive liquid waste generated during decommissioning operations will be a result of normal system processing and evolutions, such as, draining the reactor tank and cooling system and decontamination of areas. The radioactive liquid waste will be processed through a system that is designed to keep releases of radioactive materials to unrestricted areas below regulatory limits. Some very low level radioactive liquid waste will continue to be generated as a result of water purification systems operations required to meet current facility technical 132

specifications. During the early stages of the decontamination and decommissioning process, the currently installed radioactive waste will require dismantling. Any additional liquid radioactive waste processing requirements will be undertaken utilizing contractor supplied (temporary) mobile liquid waste processing equipment.

It is the goal of the Omaha VAMC management that no liquid waste will be shipped from the AJBRF. Liquid waste will be processed on site, using either the existing demineralizer system or contractor supplied system to reduce the activity of the liquid waste below regulatory limits. Subsequently, liquid waste will be stabilized and solidified and shipped as solid waste, as indicated above.

Contractors radioactive waste processing equipment and operations will be reviewed in accordance with the requirements of 10 CFR 50.59 and the AJBRF technical specifications. Formally established and controlled implementation procedures will be developed, as necessary, to ensure liquid radioactive waste processing controls will be maintained.

12.2.7 Airborne Radioactive Waste Processing Airborne radioactive waste processing equipment consists of ventilation fans, ducting, dampers, louvers, filters, stack and controls. The purpose of the airborne waste processing system is to provide for the monitoring and control of airborne radioactive releases and provide sufficient ventilation to minimize airborne contamination within the reactor facility.

Exhaust air from the reactor facility will be discharged through temporarily installed HEPA systems installed outside of the facility. Monitoring equipment continuously drawing a representative sample from the discharge side of the HEPAs will monitor radioactive effluents. Procedures will be established and controlled that provide for sampling, measuring, and reporting radioactive airborne releases in order to ensure that airborne releases are monitored and maintained within the limits of the AJBRF technical specifications.

Airbome radioactive waste generated during the decontamination and decommissioning will primarily consist of particulates originating from various decontamination and dismantling activities. Dismantlement activities will be planned and performed to ensure that airborne releases are minimized and monitored to the maximum extent practicable by implementing the following considerations during detailed planning of decommissioning activities:

• Maintaining an airborne waste processing system with a temporary/portable system containing HEPA filters;

• Using local HEPA filtration systems when specific localized activities could result in the release of significant radioactive particulates; 133

• Using the additional general area HEPA filtration when local HEPA filters are not practicable or their use alone could result in the release of significant radioactive particulates;

• Establishing controls to require local monitoring at the point of release of the temporary ventilation; and

• Establishing procedures for the analysis of airborne effluents through all significant pathways utilized during decommissioning. These procedures will address how the radioactive effluent releases are in compliance with 10 CFR 50 limits and 10 CFR 20 concentration limits.

12.2.8 Mixed Low Level Radioactive/Hazardous (Mixed Waste)

The handling and disposal of hazardous materials and hazardous wastes will be controlled through the Omaha VAMC hazardous waste management program.

Potential mixed waste anticipated to be generated during decommissioning includes the paint applied to the reactor walls during the construction of the AJBRF, and the floor tiles in the reactor pit/control room. The lead and asbestos containing materials will be classified and, if appropriate, disposed of at an Environmental Protection Agency (EPA)/NRC authorized facility.

No chemicals or other substances are anticipated to be used during decommissioning operations that could generate a mixed waste. If mixed wastes are generated, they will be managed according to EPA regulations, issued under Subtitle C of the Resource Conservation and Recovery Act (RCRA) (Act 64)

Rules and according to Nebraska Department of Environmental Quality Title 128 requirements to the extent they are not inconsistent with federal handling, storage and transportation requirements.

I Mixed wastes from the AJBRF will be transported and shipped by licensed transporters to authorized and licensed processing and disposal facilities.

Vendors will be selected based upon evaluations of their proposed processes to render the mixed wastes non-hazardous.

12.2.9 Radioactive Waste Minimization Radioactive waste minimization will be practiced at the AJBRF throughout the decommissioning process. This will be achieved by management focus, worker training, and written procedures to ensure volume reduction is considered in tasks. Tools used in the AJBRF will be controlled and re-used to minimize the number of contaminated tools, and all equipment and materials provided by outside vendors shall be required to be designed to facilitate decontamination.

An aggressive program of minimizing packaging and other unnecessary materials inside the controlled areas of the AJBRF will be implemented.

134

All workers that frequent the radiologically controlled area of the reactor facility will receive Basic Radiation Worker Training. This training will cover the purpose, effect and benefits from an effective volume reduction program.

Off-site vendors for radioactive waste processing will be utilized, as necessary, during the various stages of the decommissioning process. These vendors typically utilize techniques that ensure that maximum efficiency for each package of radioactive waste is obtained. These techniques include incineration of dry active waste, super compaction, melting, and decontamination for free release.

Packaging of radioactive waste during the dismantlement process will be performed in accordance with written procedures. A primary objective of these procedures is to ensure that voids in packages are minimized. Void minimization helps to ensure that the radioactive waste volume generated has been reduced to the maximum extent practicable. Such volume reduction minimizes the project costs, disposal and transportation risk.

12.3 Non-radioactive Hazardous Waste Management During the operational phases, the reactor typically had been classified as a Conditionally Exempt hazardous waste generator. Examples of routinely handled hazardous waste generated during operation include lighting waste and chemicals. Non-hazardous waste produced during routine operations includes items such as lubricating oils and trash. A general cleanup at the AJBRF has resulted in the removal and disposal of all non-radioactive chemicals and other potentially hazardous waste.

12.3.1 Waste Management Statement The Omaha VAMC is committed to the safe decommissioning of the AJBRF. The primary objective of the Omaha VAMC's hazardous waste management plan is to protect workers, the public, and the environment from the potential effects of hazardous wastes. The Omaha VAMC is committed to strict compliance with all federal and state hazardous waste handling and disposal requirements.

12.3.2 Hazardous Material Management The OSHA Hazard Communication requires the Omaha VAMC (29 CFR 1910.1200) to provide information to its employees and contractors concerning the hazardous substances to which they may be exposed. The Omaha VAMC's Right to Know Program was implemented to meet these requirements. General employee training is provided to apprise employees and contractors of the types and identification of hazardous materials used at the AJBRF.

135

12.3.3 General Waste Management Program A formally established Waste Management Program will be applied to decommissioning wastes. Programs for the management of waste and pollution prevention will be defined through the Omaha VAMC safety manual and AJBRF radiation protection procedures.

The requirements for the disposition of equipment and materials will depend on whether the material or equipment is categorized as a product or as a waste. The categorization of the waste will be based on: (1) Federal RCRA, (2) State of Nebraska Title 128 Requirements, (3) its physical and chemical properties, and (4) the disposition of the material.

Materials and equipment that can be used without reclamation are typically classified as products. Disposition of products will be coordinated through the Omaha VAMC Acquisition & Materiels Management Service.

During the dismantlement phase, a waste handling storage area will be established to facilitate the inventory, collection, categorization, and disposition of waste materials in containers. The area will provide adequate containment and segregation of potentially incompatible or reactive wastes, and will minimize potential for environmental release.

Chemicals used during decommissioning will be evaluated for hazardous constituents or properties using RCRA criteria and the chemical's Material Safety Data Sheets (MSDS). Decontamination agents expected to be used during decommissioning activities include detergents and solvents. Detergents and water-based solvents that would generate a non-hazardous waste will generally be used for cleaning.

Efforts will be made to minimize the production of wastes. All chemicals and materials used during the decommissioning will be subjected to a chemical control review by the Omaha VAMC Industrial Safety Manager to determine if a non-hazardous or a less toxic chemical may be substituted to prevent the generation of mixed wastes. In the event that hazardous chemicals or materials must be used, waste minimization techniques will be applied during usage. Steps will be taken to ensure that if a potentially hazardous material must be used, controls are in place to ensure these materials are not inadvertently contaminated with radioactivity. If any hazardous material becomes radioactively contaminated, it will be considered a mixed waste, subject to applicable NRC, EPA and state regulations. Any such mixed waste generated will be managed according to Subtitle C of RCRA and applicable state and local permitting and operating requirements.

136

12.3.4 Specific Waste Management Based on waste stream inventory data, wastes that have the potential to be hazardous will be classified through MSDS information or analysis. A sampling, analysis and composite method will be utilized to properly classify and group similar wastes. Following classification, appropriate regulatory waste disposition options will be evaluated and selected. Selection of disposition methods, in order of priority, will focus on: (1) reuse/recycle, (2) on-site elementary neutralization of acids or bases, and (3) off-site treatment.

12.3.5 Mercury Instruments and switches which contain mercury will be consolidated and either drained or the entire assembly shipped off-site to a licensed mercury reclamation or disposal facility. Mercury drained from equipment will be reclaimed or processed by an authorized contractor.

12.3.6 Lead-Based Paints Historically, lead-based paints may have been used to coat steel components, concrete structures and the reactor liner. During the operating life of the reactor facility, some lead-based paints may have been covered with several coats or types of paint. In other cases, non-lead-based paint surfaces may have been coated or touched up with a lead-based paint. Lead-based paint identification and removal process controls will be implemented, in accordance with a controlled lead paint removal program to ensure proper handling of surfaces painted with lead-based materials and disposal of wastes. Metal components coated with lead-based paint that meets the radioactive release criteria will be sent to a licensed scrap metal facility.

12.3.7 Lead Lead from various locations of the reactor facility, mostly used for shielding in the form of casks, pigs and bricks that can be decontaminated to meet the radioactive release criteria will be sent off-site for scrap metal recycling per Section 12.3.9. Lead containing material that cannot meet the radioactive release criteria and/or the scrap metal criteria will be characterized and disposed of at a licensed facility.

12.3.8 Insulation/Asbestos-Bearing Type Materials Asbestos-type material may exist in floor tiles both in radioactively and non-radioactively contaminated areas of the reactor facility. This will be confirmed prior to commencement of decommissioning operations. Radioactively-137

contaminated asbestos materials will be collected and disposed of in accordance with Section 12.2.2.

Asbestos removal work will be performed using appropriately trained personnel.

Asbestos will be packaged for shipment and disposed of at an authorized disposal site. Asbestos handling and disposal regulations will be enforced.

Generally, asbestos-containing materials will be handled in, such a manner as to minimize air emissions, collected in plastic bags and labeled "Asbestos-Containing Waste." Disposal of radioactively clean asbestos will be prearranged at a licensed, Type II solid waste disposal landfill.

12.3.9 Scrap Metal During dismantlement, non-radioactive structural metal components, plates, piping and wire cables will be removed and sent off-site for scrap metal recycling.

Radioactive contaminated metals that can economically be decontaminated to meet the radioactive release criteria will be decontaminated and sent to scrap metal recycling facilities. Metals that cannot be decontaminated will be characterized and disposed as radioactive waste.

12.3.10 Miscellaneous Non-Hazardous Solid Waste Decommissioning and dismantlement will require the disposal of system and building wastes. These wastes will include materials that were never radiologically contaminated or otherwise meet the radiological-release criteria, and that are not classified as hazardous waste. Non-radioactive, non-hazardous wastes are expected to include:

• Duct work and associated equipment including, ducts, fans, filters and supports;

• Electrical systems and equipment, such as cables, conduit and motors;

• Concrete flooring and bituminous pavement (asphalt-type); and

• Office furniture and fixtures.

The priority options for these materials are: (1) re-use within the Omaha VAMC, (2) recycle, and (3) disposal at a licensed Type II landfill.

12.3.11 Non-Hazardous Liquid Wastes Following classification as non-hazardous waste, contained liquids will be categorized to facilitate recycling, re-use, and/or disposal. Liquids transported off-site will be manifested and handled by approved disposal/recycling facilities.

138

12.4 Potential Environmental Response In the event that a hazardous substance release occurs or is discovered prior to or during decommissioning, arrangements will be made to continue or initiate remedial actions prior to site closure in conjunction with the Omaha Fire Department/HAZMAT personnel.

The existing procedures for spill control and mitigation will be enhanced to address the additional handling of hazardous materials during the AJBRF decommissioning. An AJBRF-specific Spill Control and Countermeasures and Pollution Incident and Prevention Plan will be established in coordination with Omaha HAZMAT/Fire Department personnel and remain in effect until such time as all of the materials in the plan, or additional polluting materials resulting from decommissioning/decontamination activities have been removed from site. If a spill occurs during the decommissioning process, it will be handled according to the plan and written implementation procedures.

12.5 Spent Fuel Management The AJBRF does not contain any nuclear fuel. The reactor was de-fueled in June of 2002. Fuel, including fission chambers, was transferred to the USGS research reactor facility in Denver, Colorado.

139

13 Quality Assurance Program The Omaha VAMC D&D team is committed to comprehensive and effective Quality Assurance (QA) program as an integral part of the D&D effort. QA will provide a systematic approach to the execution and review of key activities and will assist in providing assurance that program activities are conducted in a manner that complies with established policies, procedures and recognized good practices. Objectives of the Omaha VAMC's decontamination and decommissioning effort will include:

1. Prevention of uncontrolled release of radioactive materials outside of the reactor facility site;

2. Limiting radiation exposure to workers and to the public at large to levels below those stipulated in 10 CFR 20;

3. Package and ship radioactive and/or hazardous wastes or materials and to characterize and measure waste for appropriate disposal within the guidelines provided in 10 CFR 61 and 71, and 49 CFR 172 and 173;

4. Controlling work practices and procedures to comply with federal, state, and local requirements;

5. Preventing the unnecessary spread of radioactive and/or hazardous contamination to uncontaminated areas; and

6. Conducting work in a safe manner.

To implement QA, the Omaha VAMC will exclusively utilize contractors with approved QA programs (per 10 CFR 50, Appendix B) and with demonstrated experience in D&D projects. Such programs will require, at a minimum, the following activities to be documented:

1. Contractor Authority and Responsibility: Including, but not limited to, written definitions of authority, duties, and responsibilities of contractor managerial, operation, and safety personnel; a defined organizational structure; and assigned responsibility for review and approval of plans, designs, procedures, data, and reports.

2. Personnel Training: Including, but not limited to, training program to provide staff trained and qualified in principles and techniques of jobs assigned, aware of the nature and goals of the QA and demonstrated proficiency maintained by retraining and/or periodic performance reviews.

3. Procedures: Written procedures for decommissioning activities (including, but not limited to, operational activities, surveys, sampling activities, sample chain of custody, calibration of instruments, and equipment maintenance and 140

calibration) that are prepared, reviewed, and approved by knowledgeable/responsible persons.

4. Documentation and Data Management: Documentation of activities performed and to track and control tasks in progress from commencement to conclusion.

5. Corrective Action Process: Process to investigate and correct recognized deficiencies and document corrective actions.

The Omaha VAMC D&D team will also implement policies and procedures described in Sections 13.1 through 13.5 of this document.

13.1 The Decommissioning and Decontamination Project QA Organization The D&D Project QA organization will develop and deploy planned systematic actions to provide acceptable confidence in D&D program results. The foundation of the organization is a management structure that relies upon experienced and knowledgeable personnel including the Decommissioning Contractor Manager, members of the Safety and Oversight Team comprised of the Quality Oversight Team, the Safety Manager, and the RSC. Section 9 of this plan outlines specific roles and responsibilities relevant to quality assurance.

The D&D Project QA organization will be required to apply site-specific knowledge and requirements to appropriately execute the quality assurance programs of contractors with 10 CFR 50, Appendix B compliant programs.

Those programs will be reviewed to best determine how to address the eighteen criteria of 10 CFR 50, Appendix B within the specific AJBRF environment.

Contractor performance will be assessed by periodic audits by the Safety and Oversight Team and will include, but not be limited to, a review of the documented work plans, processes, and procedures that the contractor uses or intends to use in conducting a specific activity, a post performance determination of the adherence to the documented work plans, processes, and procedures, and a review of the timeliness and safety of the work performed (e.g., accident analysis).

Ultimate responsibility for safely conducting the D&D and the implementation of and adherence to the QA plan is the responsibility of the ACOS/R and the Quality Oversight Team. To ensure ongoing QA, all program management personnel will be required to review the overall QA program (as outlined herein) and will document any deviations from the QA plan either by Omaha VAMC program management or contractor personnel.

A high-level organization chart of the Omaha VAMC QA D&D management structure is outlined in Figure 9.1.

141

13.2 The Omaha VAMC Quality Assurance Program The Decommissioning Contractor Manager, in conjunction with Safety and Oversight Team, will institute the following processes and procedures to ensure minimal exposure to radioactive and/or hazardous materials.

1. All work done on the decommissioning and decontamination of the reactor and surrounding environment will be contracted only to organizations that have a QA program that has a 10 CFR 50, Appendix B, equivalent program.

2. Contractors' QA programs will be reviewed in advance by the RSC manager(s), in conjunction with the Safety and Oversight Team to ensure that the programs are properly documented and controlled and are compliant with required regulatory and licensing requirements.

3. All work plans, programs, and procedures to be utilized by contractors will be reviewed in advance by the Omaha VAMC senior management, the ACOS/R, RSC, as well as the Safety and Oversight Team to ensure that work plans, programs, and procedures implement the approved QA program

4. Contractors will be required to submit detailed reports of activities on a periodic basis. Such reports will be reviewed by the ACOS/R and the Safety and Oversight Team

5. Any substantive changes that contractors propose to make (beyond editorial changes) to their QA programs will be reviewed in advance by the Safety and Oversight Team and the RSC.

13.3 Document Control The ACOS/R, in conjunction with the RSC and the QOT, will institute the following processes and procedures to ensure appropriate control of QA documentation.

1. All work done on the decommissioning and decontamination of the reactor and surrounding environment will be contracted only to organizations that have a QA program that is 10 CFR 50, Appendix B compliant which specifies document control standards.

2. As part of the contract evaluation procedure, the ACOS/R, in conjunction with the RSC and the QOT, will review the QA document control procedures of the proposed contractors.

3. The ACOS/R, in conjunction with the RSC and the QOT, will require that all contractors utilize a documented policy and supporting procedures to ensure that all QA-related documentation developed, issued, and revised is conducted in a controlled manner.

142

The purpose of the review of the licensee's or responsible party's QA program is to determine that adequate control is maintained of document issuance, changes and storage.

13.4 Control of Measuring and Test Equipment (M&TE)

The Omaha VAMC as part of the commitment to minimize possible exposure to radioactive and/or hazardous materials, the Omaha VAMC D&D team will require that any contractor utilized will have a test and measurement equipment calibration program in place that meets the requirements stipulated under 10 CFR 30.36(g)(4)(ii), 40.42(g)(4)(ii), 40.28(b)(3), 70.22(f), 70.38(g)(4)(ii), and 72.54(g)(6). To ensure compliance, the ACOS/R, via the RSC and the QOT, will institute the following processes and procedures.

1. As part of the procurement process governing selection of contractors, all contractors will be required to provide a detailed summary of the specific M&TE proposed for use.

2. The contractors will be required to provide a documented calibration schedule (daily, weekly, monthly, annually, as appropriate) for the proposed test equipment. The schedule will be reviewed by the ACOS/R and QOT, and approved by the RSC.

3. The contractors will provide, as part of their management reporting, a detailed listing of the calibration of testing equipment. Such a report will be required to be produced on a monthly basis at a minimum. These reports will be provided to the ACOS/R who, in conjunction with the QOT, will review and ensure that calibration and calibration checks are performed.

4. The contractors will be required to maintain a controlled inventory of all M&TE used (including documentation of the manufacturer, model, and serial number, as appropriate). The contractor will be required to cross-reference this listing with the calibration and calibration check report produced.

13.5 Corrective Action The key responsibility of the ACOS/R is to ensure that contractors perform their duties according to the requirements of the contract and in a manner that is safe and that will minimize possible exposure to radioactive and/or hazardous materials. The Omaha VAMC staff will meet regularly with the contractors and will institute the following processes and procedures.

1. As a standard part of the regular meetings with the contractors, the ACOS/R will require the contractors to identify any possible QA issues. If issues are identified, the ACOS/R will require that the contractors determine appropriate corrective measures.

143

2. Upon determination of corrective measures presented by the contractors, the ACOS/R, in conjunction with the RSC and the QOT, will review and approve proposed corrective actions upon concurrence of the adequacy of the corrective action.

3. Each contractor will be required to maintain a record of issues and corrective actions taken. In addition, the contractor will be required to report issues and corrective actions as part of regular management reporting to the ACOS/R and support staff.

13.6 Records Management The Omaha VAMC D&D team is committed to developing and maintaining a comprehensive record of all QA actions, documents, and policies and procedures. To that end, the ACOS/R, in conjunction with the RSC and the QOT, will institute the following policies and procedures.

1. Fundamental QA record management will be the responsibility of the appropriate contractor. All contractors will be required to have in place a documented records management methodology. Prior to commencement of any project phase, the ACOS/R, in conjunction with the RSC and the QOT, will review the contractors documented records management methodology for adequacy.

2. At the completion of the D&D project, the contractor will be required to present the ACOS/R with a complete inventory and duplicate copies of all QA-related documents related to the work completed.

3. Upon receipt and review of the documents referenced in the procedure noted above (2), the ACOS/R will transfer the documents to a location physically separated from the project environment. The location will be a secure location.

13.7 Audits and Surveillances Exclusively, contractors with a 10 CFR 50, Appendix B program compliant QA program will conduct D&D activities. The Omaha VAMC D&D team, represented on a day-to-day basis by the Project Controls and Oversight Group, will take an active role in monitoring the contractors. Subsections 13.1 through 13.6 detail the actions that will be taken by the ACOS/R, including, but not limited to, detailed management reporting by contractors, review and approval of substantive changes of contractors' QA programs, and ongoing identification and resolution of QA issues. To further ensure that the contractors perform as required, the ACOS/R, in conjunction with the RSC and the QOT, will institute the following audit and surveillance processes and procedures.

144

. The QOT and the RSC will review and audit the QA records, policies and procedures on a regular basis.

2. The QOT will audit, through inspection and observation, contractors' work conduct to ensure compliance with the documented QA policies and procedures.

3. A record of those audits will be captured, documented, and will become part of the permanent records related to the project and will be managed through the documents management procedures outlined in Section 13.6.

4. Any issues resulting from the audit will be communicated in writing to the contractor and will include a requirement for the contractor to present a written resolution to the issue within three business days after issue presentation.

5. Contractors will be required to report on the status of issues and the correction of issues through the regular management-reporting program.

6. Contractors will be required to certify, in writing, that the issues have been resolved and provide adequate documentation as to the resolution in a timely manner. All documents related to the audit results and resolution will become part of the permanent records and will be managed through the documents management procedures outlined in Section 13.6.

7. In appropriate situations, the ACOS/R will ensure through inspection and observation by the QOT that contractors' resolution actions are in place and effective. Such a review will take place no later than ten business days after presentation of the certification outlined in the above noted procedure (6).

145

14 Facility Radiation Surveys 14.1 Release Criteria The Federal regulation on License Termination, 10 CFR 20.1402, defines the radiological limits for release of a licensed site for unrestricted use. Acceptable criteria for unrestricted release conditions are set at a maximum TEDE of 25 mrem per year from residual radioactivity above background, including that from ground water sources of drinking water, and that the residual radioactivity has been reduced to levels that are ALARA.

MARSSIM provides options for designing final surveys for sites with varying characteristics applicable to surface, soil, and building characteristics. Section 2.6 of MARISSM discusses these options or alternatives in detail. Specifically Sub-Section 2.6.1 discusses the approach of using statistics to provide a quantitative estimate of the probability that the release criterion is not exceeded at the site versus a simpler methodology using a comparison to an investigation level to evaluate the presence of small areas of elevated activity in place of complicated statistical tests. Furthermore, it is recommended that the use of simple comparisons to each measurement result to the DCGL, to demonstrate that all the measurement results are below the release criterion, may be more effective than statistical tests for the overall demonstration of compliance with the regulation provided an adequate number of measurements are performed.

The Omaha VAMC plans to implement the alternate methodology described in MARSSIM Section 2.6.1, which provides guidance for direct comparison of each measurement result to the DCGLs. The Omaha VAMC believes that this recommended methodology is practical and warranted at a facility, such as the AJBRF, where low levels of radioactivity exists, with the exception of the reactor pool and components and facility drain lines, which will be removed from the facility. The overall result from implementing this ALARA methodology is reduced risk to future facility occupants following license termination.

With regard to evaluating environmental data, MARSSIM recommends the utilization of nonparametric statistical tests methodology. The AJBRF staff intends to utilize the applicable guidance contained in NUREG-1 505, 'A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys", for evaluation of environmental data in meeting release criteria.

DCGLs were developed as part of the site characterization effort and are documented in the site characterization survey report (Attachment A). These 146

DCGLs, in combination with the existing AJBRF material release limits, will be utilized as the regulatory basis for release of the site for unrestricted use.

14.2 Site Characterization Reports 14.2.1 General Overview The facility characterization survey is an important part of the overall AJBRF decommissioning planning and license termination process. The site characterization survey can be the most comprehensive of all the survey types and typically generates the most useful data for determining the extent and methodology for decommissioning.

Understanding of the importance to properly perform a site-specific characterization at the AJBRF, the VA issued a contract to GTS Duratek, in September of 2002 for the performance of a comprehensive site characterization survey. The procurement specifications required that the contractor utilize NUREG-1 575, MARSSIM as the basis and guidance document for planning and executing the site characterization survey. The characterization of the AJBRF was conducted during the period October - November 2002. The site characterization report and supporting details are appended to this decommissioning plan as Attachment A.

The site characterization process included the following activities:

• Background radiation study of surfaces of similar materials to those of the AJBRF;

• Radiological characterization of the areas outside and adjacent to the AJBRF;

• Radiological characterization of the rooms, structures and components;

• Radiological characterization of laboratory fume exhaust hoods, facility ventilation system and the areas near exhaust points;

• Neutron activation analysis of areas surrounding and outside the reactor pool;

• Vertical core boring of subsurface areas in proximity of the reactor pit and sampling for activation and contamination;

• Performing hazardous sampling and analysis for bulk asbestos and lead paint; and

• Identification and preparation of estimated quantities of radioactive and hazardous waste including defined volumes, linear feet of contaminated piping, imbedded piping, and determination of alternatives for disposition of all materials.

147

The site characterization report with appendices (Attachment A) contains detailed information relative to the site characterization report design and the results of the survey. The site characterization report includes descriptions of survey measurements, instrumentation for direct measurements, analytical procedures, specific survey results, maps and/or drawings of the site and areas surveyed within and adjacent to the AJBRF, and determination of impacted and non-impacted areas of the facility.

The Omaha VAMC believes that the characterization that was performed is adequate based on the following factors and their correlation with the survey results:

• The AJBRF is an extremely low power reactor that was under utilized, especially over the last ten years and had minimal lifetime fuel burn-up;

• Historically, environmental monitoring results indicated non-detectable releases of radioactive material from reactor operation;

• Direct radiation levels were generally at background levels;

• A stringent vendor QA program and independent oversight was applied to the process;

• Industry-accepted instrumentation and implementation procedures were utilized for the characterization; and

• An experienced vendor that has successfully completed similar characterizations at other NRC licensed sites was utilized.

All areas of the AJBRF were found to be accessible to the extent necessary for a comprehensive site characterization effort. During the characterization survey, direct radiation readings were taken within the reactor tank. Ventilation systems were opened and surveyed. Facility drain lines were measured internally, to the extent possible with detection probes. Also, core borings around the reactor pit and outside areas adjacent to the facility were conducted. It should be noted that drain lines were accessed to the degree that elevated contamination levels were found within two feet of the opening of the drains.

Based on a review of the characterization survey results the Omaha VAMC believes that the levels of radioactivity at the AJBRF have been adequately characterized and that it is most unlikely that significant quantities of radioactive have gone undetected.

During the course of the site characterization, a sample was collected from the demineralizer resin. A 10 CFR 61 analysis was performed by an independent laboratory on the resin sample. The results of the analysis, which is believed to be a conservative representative of the radioisotope mix existing at the AJBRF, was used as the basis for the calculation of the DCGLs to be used for the finalltermination survey in support of unrestricted release. The results and isotopic analysis for this sample is contained in the characterization report 148

(Attachment A). The sample results clearly identify individual radionuclides along with their respective ratios.

14.3 Remediation Action Support Surveys 14.3.1 General The Omaha VAMC intends to implement Remedial Action Support Survey (RASS) methodology using the applicable guidance contained in MARSSIM Sub-Section 5.4. The objective of the RASS will be to:

• Perform surveys as necessary to support decontamination and decommissioning activities;

• Determine when a site or survey unit has been decontaminated below DCGL levels and is ready for the final status survey; and

• Provide updated estimates of site-specific parameters to use for planning the final status survey.

In addition, as discussed in Section 10 of this decommissioning plan, routine and operational surveys for surface contamination, radiation, and airborne radioactivity will be performed as necessary to measure the status of radioactive contaminants remaining on structures and equipment during the decommissioning effort.

14.3.2 Survey Design A critical objective of the RASS is to detect the presence of radioactive contaminants at or below the applied DCGL criteria. The goal will be to set the RASS significantly below this defined DCGL. In order to meet this survey goal the selection of instrumentation and techniques will be based on the detection capabilities for the known or suspected radioactive contaminants and levels of contamination to be measured.

14.3.3 Conduct of Surveys Radiological monitoring instrumentation and operating procedures will be selected based on specific detection capabilities commensurate with the expected radioactive contaminant profile. Survey methodology for the identification of specific radio-isotopes will typically include a variation of surface scans complimented by taking direct measurements.

149

14.3.4 Evaluation of Survey Results Consistency in the use of standard units for data reduction is an important part of the successful evaluation of the decommissioning process. Procedures will be implemented to ensure that radioactivity measurements obtained from scanning efforts and laboratory analysis are converted and calculated in a consistent and reliable manner. Stringent analytical techniques are vital to the decommissioning process in determining whether or not more remediation is required or that the radioactive contaminant is at a level below the applied DCGL.

The Omaha VAMC intends to utilize similar instrumentation for the decontamination and decommissioning activities as was used for the site characterization survey, in terms of instrument detection capabilities, operational characteristics, mobility, detector sensitivity and reliability.

14.4 Final Status Survey Design The purpose of the final survey will be to demonstrate that the AJBRF has been decommissioned to the extent necessary to meet the unrestricted use release requirements as specified in 10 CFR 20.1402 'Radiological Criteria For Unrestricted Use". The basis for developing the final survey plan will consist of applicable guidance contained in NUREG-1575, MARSSIM and the guidance contained in NUREG-1 505, "A Non-Parametric Statistical Methodology for the Design and Analysis of Final Status Surveys".

The final survey plan will be designed and implemented in such a manner that the requirements of the decontamination and decommissioning plan and associated implementation procedures are met, thus allowing for license termination. The final survey plan design will be based on the assumption that the remaining activity is essentially normally distributed relative to both the interior and exterior areas of the AJBRF. If the results of the survey demonstrate that this assumption is not true, then the final survey implementation procedures will be revised to accommodate the different conditions.

14.4.1 Survey Design Data Quality Objectives (DQO)

The DQO process will consist of a series of planning steps, using a graded approach, to ensure that the types, quantity, and quality of radiological data used in decision-making are appropriate for the intended application. DQOs are qualitative and quantitative statements that clarify the process objective, define the most appropriate data to collect, determine the most appropriate conditions for collecting the data and specify acceptable levels of decision errors that will be 150

used to establish the quantity and quality of data needed to support the decision.

The DQOs intended to be used for the AJBRF are as follows:

• Proper selection and independent verification of survey unit classification;

• Collection of sufficient data of high quality to ensure that a comparison can be made with the release criteria for each survey unit to determine if residual radioactivity in each unit has been reduced to a level below the release criteria; and

• Ensuring that the potential radiological risk from the AJBRF as a whole is below that represented by the dose limit release criteria.

14.4.2 Survey Area Classification The final survey classification for the AJBRF will be subdivided into three distinct classes:

• Class I areas: Areas that have, or had prior to remediation, a potential for radioactive contamination or known contamination, in excess of DCGLw.

• Class 2 areas: Areas that have, or had prior to remediation, a potential for radioactive contamination or known contamination, but are not expected to exceed the DCGLw.

• Class 3 areas: Any impacted areas that are not expected to contain any residual radioactivity, or are expected to contain levels of residual radioactivity at a small fraction of the DCGLw.

Class I areas will receive the highest degree of survey efforts during the final survey.

The site characterization survey (Attachment A) serves as the basis for determining the classification of areas for the final status survey.

14.4.3 Survey Unit Designation The designation of survey units for the final status survey will be very similar to those areas identified and documented in the attached site characterization report. Maps and drawings incorporated in that report shall be utilized for the development of the final status survey plan. As a minimum, a survey documentation package will be established for each survey unit defined in the site characterization report. Each survey package will include, as a minimum:

• General work instructions that will define the measurements, instrumentation to be used and instructions for conduct of the survey; 151

• Survey location designations, results and comments regarding conditions encountered; and

• Map(s) of the area to be surveyed.

14.4.4 Survey Locations and Survey Points Survey locations will be clearly identified to allow for reproducibility of measurements and to provide a method of referencing survey results to survey units.

14.5 Conduct of the Final Status Survey 14.5.1 Instrumentation and Sampling Radiation monitoring instrumentation to be utilized in the final status survey will have the necessary sensitivities and capabilities to detect the radiation of interest and at levels that meet or exceed applicable legal and administrative criteria.

The instrumentation to be utilized will be calibrated, operated and maintained in accordance with approved AJBRF radiation protection procedures.

A list of the types of radiation monitoring instrumentation to be utilized for the final survey effort is included in Section 10 to this D&D plan. In addition, the types and capabilities of the instrumentation listed in the site characterization report (Attachment A) will be considered for use based on equipment availability.

MDAs for all instrumentation used for the final survey effort will be established in accordance approved instrument calibration and operating procedures.

14.5.2 Background Determination During the site characterization survey, background levels were established and utilized as part of the site assessment. The background will be re-validated as a prerequisite to commencement of the final status survey.

14.5.3 Measurement Frequency The estimated sample sizes for each survey area will be determined using the guidance in Table 5.3 of MARSSIM. However, if warranted, a larger number of samples will be added to survey units on a case-by-case basis based on the judgment of the survey contractor, the RSO and the RSC.

152

14.5.4 Sample Collection Samples obtained during the final status survey will be collected, controlled and stored in accordance with approved AJBRF radiation procedures. Rigorous chain-of-custody practices will be implemented to ensure control and the reproducibility of samples.

14.5.5 Quality AssurancelQuality Control The quality assurance plan described in Section 13 of this decommissioning plan will be implemented, as applicable, to ensure that the quality and regulatory requirements for conduct of a final survey are met. Activities affecting quality will have controls established to ensure that the appropriate equipment, required environmental conditions, and prerequisites for the given activity have been met.

Specific quality assurance criteria applicable to the final status survey will identified in a stand-alone quality assurance plan. It is expected that the vendor contracted to perform the final status survey will provide this quality assurance document, which will be approved by the AJBRF RSC prior to implementation.

Quality control will be implemented through the RSC approved procedures. This will be a multi-faceted program to ensure the quality and accuracy of the survey data. This program will include quality control over field and laboratory instrumentation, control of samples, sample analysis, use of radioactive reference sources, and data processing, including management and control.

14.6 Final Status Survey Report Following completion of the final status survey, a final status survey report will be developed and submitted to the NRC. The report will contain information needed by the NRC to make a decision concerning termination of the AJBRF license and authorization of the site for unrestricted use. The report will be prepared using the applicable guidance contained in Section 8.6 of MARSSIM.

153

15 Financial Assurance The Omaha VAMC in coordination with the Director of Finance, Office of Research and Development, Department of Veterans Affairs is responsible for developing and submitting all budgeting and legislative requests necessary to operate, maintain, and ensure the ultimate proper disposition of the AJBRF. In accordance with 10 CFR 50.75(e)(1 )(iv), the Department of Veterans Affairs has requested and received funds consistent with a $5,200,000 cost estimate for the planned decommissioning activities.

Table 15.1, Decommissioning Project Cost Estimate

.- Activity

'Estimated Cost in Duration in

$K'

-'Months':.

Decommissioning 1,200 10 Design and Planning Decommissioning 3,200 6

Final Survey &

800 4

License Termination Total 5,200 20 Should additional funds, above and beyond those designated for the decommissioning be required, the Department of Veterans Affairs will engage in the required budget processes necessary to ensure that required funds for decommissioning are provided consistent with the work plan provided in Section 8.

154

16 Restricted Use/Alternate Criteria The objective of the AJBRF decommissioning is the regulatory release of the reactor site and adjacent areas to unrestricted use conditions. For additional information relating to the decommissioning option selected, refer to Section 6 of this plan.

155

17 References 17.1 Code of Federal Regulation (CFR) Parts

• 10 CFR 19, Notices, Instructions and Reports to Workers: Inspection and Investigations

• 10 CFR 20, Standards for Protection Against Radiation

• 10 CFR 30, Rules of General Applicability to Domestic Licensing of Byproduct Material

• 10 CFR 40, Domestic Licensing of Source Material

• 10 CFR 50, Domestic Licensing of Production and Utilization Facilities

• 10 CFR 61, Licensing Requirements for Land Disposal of Radioactive Waste

• 10 CFR 70, Domestic Licensing of Special Nuclear Material

• 10 CFR 71, Packaging and Transportation of Radioactive Material

• 10 CFR 72, Licensing Requirements For The Independent Storage Of Spent Nuclear Fuel and High-Level Radioactive Waste

• 10 CFR 100, Reactor Site Criteria

• 29 CFR 1910, Occupational Safety and Health Standards

• 29 CFR 1926, Safety and Health Regulations for Construction

• 40 CFR 24, Protection of Environment

• 49 CFR 172, Shippers: General Requirements for Shipments and Packagings 17.2 NUREGs

• NUREG-0041, Manual of Respiratory Protection Against Airbome Radioactive Material

• NUREG-1 505, A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys

• NUREG-1 575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM)

• NUREG-1727, ALARA Analyses 17.3 Other Referenced Materials

• AJBRF Initial Operating License (R-57, issued 6/26/1959)

• Amendment 9 (issued 4/12/1991) 156

• Amendment 11 (issued 8/5/2002)

• AJBRF Safety Analysis Report (SAR), August 2002

• American Conference of Government and Industrial Hygienists (ACGIH)

• American National Standards Institute (ANSI) N323 - 1978, Radiation Protection Instrumentation Test and Calibration

• Atomic Energy Act of 1954

• Construction Permit CPRR-36

• Duratek Site Characterization Survey Report, January 2003

• Environmental Protection Agency 52011-88-020

• National Institute for Occupational Safety and Health/Mine Safety and Health Administration (NIOSH/MSHA)

• National Institute of Standards and Technology

• National Voluntary Laboratory Accreditation Program (NVLAP)

• NRC Regulatory Guide 1.86

• NRC Regulatory Guide 8.15

• NRC Federal Guidance Report No. 11

• Reactor Safeguards Committee (RSC) minutes, 1959 to 2002 157