ML12054A119

Buffalo Materials Research Center, Decommissioning Plan

ML12054A119
Person / Time
Site:	University of Buffalo
Issue date:	02/06/2012
From:	Enercon Services
To:	University of Buffalo, NRC/FSME/DWMEP
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Download: ML12054A119 (239)

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Decommissioning Plan Buffalo Materials Research Center Prepared for:

Buffalo Material Research Center Office of Environment, Health, and Safety Services Completed by:

4490 Old William Penn Highway Murrysville, PA 15668 February 6, 2012

BMRC Decommissioning Plan Revision 0 Prepared By: ENERCON Date:

Reviewed By: Date: 2/6/2012 Robert Sanders Senior Technical Specialist Reviewed By: Date: 2/6/2012 Kurt Colborn Senior QA Specialist Approved By: Date: 2/6/2012 Dustin G. Miller, CHP DOC Project Manager

BMRC Decommissioning Plan Revision 0

SUMMARY

OF CHANGES Revisions to the Decommissioning Plan will be tracked when revisions are issued. Changed sections will be identified by special demarcation in the margin. A summary description of each revision will be noted in the following table.

Revision Date Description of Change Number 0 February 6, 2012 Initial Issue i

BMRC Decommissioning Plan Revision 0 TABLE OF CONTENTS PAGE ACRONYMS AND ABBREVIATIONS .................................................................................................... V 1.0

SUMMARY

OF DECOMMISSIONING PLAN ............................................................................ 1 1.1 Introduction............................................................................................................................ 1 1.2 Background ............................................................................................................................ 2 1.2.1 Reactor Decommissioning Overview ....................................................................... 18 1.2.2 Estimated Cost.......................................................................................................... 19 1.2.3 Availability of Funds ................................................................................................ 20 1.2.4 Program Quality Assurance...................................................................................... 20 2.0 DECOMMISSIONING ACTIVITIES........................................................................................... 23 2.1 Decommissioning Alternatives ............................................................................................ 23 2.1.1 SAFSTOR Alternative.............................................................................................. 23 2.1.2 ENTOMB Alternative .............................................................................................. 23 2.1.3 DECON-A Alternative ............................................................................................. 23 2.1.4 DECON-B Alternative.............................................................................................. 24 2.2 Facility Radiological Status ................................................................................................. 24 2.2.1 Facility Operating History ........................................................................................ 24 2.2.2 Current Radiological Status of the BMRC Facility .................................................. 25 2.2.3 Release Criteria......................................................................................................... 27 2.3 Decommissioning Tasks ...................................................................................................... 31 2.3.1 Activities and Tasks.................................................................................................. 31 2.3.2 Schedule.................................................................................................................... 36 2.4 Decommissioning Organization and Responsibilities ......................................................... 37 2.4.1 University Project Manager...................................................................................... 38 2.4.2 BMRC Director ........................................................................................................ 38 2.4.3 Radiation Safety Officer ........................................................................................... 39 2.4.4 Reactor Decommissioning Safety Committee.......................................................... 39 2.4.5 Design and Oversight Contractor ............................................................................. 39 2.4.6 Demolition Contractor.............................................................................................. 40 2.5 Training Programs ............................................................................................................... 42 2.5.1 General Site Training................................................................................................ 42 2.5.2 Radiation Worker Training....................................................................................... 42 2.6 Contractor Assistance .......................................................................................................... 44 2.7 D&D Documents and Guides .............................................................................................. 45 3.0 PROTECTION OF WORKERS AND THE PUBLIC .................................................................. 47 3.1 Radiation Protection............................................................................................................. 47 3.1.1 Ensuring ALARA Radiation Exposures................................................................... 48 3.1.2 Health Physics Program............................................................................................ 48 3.1.3 Dose Estimates ......................................................................................................... 52 3.2 Waste Management.............................................................................................................. 53 3.2.1 Fuel Removal............................................................................................................ 55 3.2.2 Radioactive Waste Processing.................................................................................. 55 3.2.3 Radioactive Waste Disposal ..................................................................................... 55 ii

BMRC Decommissioning Plan Revision 0 3.2.4 General Industrial Safety Program ........................................................................... 57 3.3 Radiological Accident Analyses .......................................................................................... 58 3.3.1 Fire............................................................................................................................ 58 3.3.2 Spill Contaminated Water......................................................................................... 59 3.3.3 Release of Airborne Contamination ......................................................................... 59 3.3.4 Transportation Accidents.......................................................................................... 60 4.0 PROPOSED FINAL STATUS SURVEY PLAN.......................................................................... 61 4.1 Survey and Sampling Approach .......................................................................................... 61 4.2 Data Quality Objectives....................................................................................................... 62 4.3 Identification and Classification of Survey Units ................................................................ 62 4.3.1 Method for Classification ......................................................................................... 62 4.3.2 BMRC MARSSIM Classifications........................................................................... 64 4.4 Data Collection .................................................................................................................... 64 4.4.1 Buildings, Equipment, and Components .................................................................. 64 4.4.2 Soils .......................................................................................................................... 64 4.4.3 Bedrock..................................................................................................................... 65 4.4.4 Bedrock Surface Scans ............................................................................................. 65 4.5 Data Evaluation.................................................................................................................... 65 4.6 Final Status Survey Report .................................................................................................. 66 5.0 TECHNICAL SPECIFICATIONS ................................................................................................ 67 6.0 PHYSICAL SECURITY PLAN .................................................................................................... 68 7.0 EMERGENCY PLAN ................................................................................................................... 69 8.0 ENVIRONMENTAL REPORT..................................................................................................... 70 9.0 CHANGES TO THE DECOMMISSIONING PLAN ................................................................... 71

10.0 REFERENCES

.............................................................................................................................. 72 APPENDIX A - ACTIVATION ANALYSIS AND COMPONENT CHARACTERIZATION APPENDIX B - SITE CHARACTERIZATION REPORT APPENDIX C - DECOMMISSIONING ENVIRONMENTAL REPORT iii

BMRC Decommissioning Plan Revision 0 LIST OF TABLES Table 1-1 Decommissioning Cost Estimate ................................................................................................ 20 Table 2-1 Estimated Volumes and Masses of Radiologically Impacted Components and Systems .......... 26 Table 2-2 NRC License Termination Screening Levels for Surface Soils ................................................. 28 Table 2-3 NRC License Termination Screening Levels for Building (Bedrock) Surfaces ......................... 29 Table 2-4 Ag-108m Surface Soil Screening Value ..................................................................................... 30 Table 2-5 Ag-108m Building Surface Screening Value ............................................................................. 30 Table 3-1 BMRC Radionuclides of Concern as determined in the Site Characterization Report .............. 47 Table 3-2 Health Physics Instrumentation .................................................................................................. 50 Table 3-3 Project Dose Estimate ................................................................................................................. 53 Table 4-1 Recommended Survey Unit Sizes .............................................................................................. 63 LIST OF FIGURES Figure 1-1 Location of BMRC on the UB South Campus ............................................................................ 1 Figure 1-2 Artists Rendering - Interior Cut Away View of the Reactor and Containment ........................ 6 Figure 1-3 Control Deck Plan View ............................................................................................................. 7 Figure 1-4 Gamma Deck Plan View ............................................................................................................. 8 Figure 1-5 Neutron Deck Plan View............................................................................................................. 9 Figure 1-6 Reactor Plan View - Control Deck Level ................................................................................. 11 Figure 1-7 Reactor Plan View - Gamma Deck Level................................................................................. 12 Figure 1-8 Reactor Plan View - Neutron Deck Level ................................................................................ 13 Figure 1-9 Reactor/Bioshield Elevation View (Northeast Direction) ......................................................... 13 Figure 1-10 Reactor/Bioshield Elevation View (Southeast Direction) ....................................................... 15 Figure 2-1 Proposed Project Schedule ........................................................................................................ 37 iv

BMRC Decommissioning Plan Revision 0 ACRONYMS AND ABBREVIATIONS ACM Asbestos Containing Material AEC Atomic Energy Commission ALARA As Low As Reasonably Achievable ALI Annual Limit on Intake ARA Airborne Radioactivity Area BMRC Buffalo Materials Research Center BSFR Bulk Survey for Release C&D Construction and Demolition CEDE Committed Effective Dose Equivalent CFR Code of Federal Regulations cm2 square centimeter D&D Decontamination and Decommissioning DAW Dry Active Waste DC Demolition Contractor DCGL Derived Concentration Guideline Levels DDE Deep Dose Equivalent DECON-A complete decontamination and structure demolition option DECON-B complete decontamination and release of the structure option DCF Dose Conversion Factor DOC Design and Oversight Contractor DOE United States Department of Energy DOT Department of Transportation DP Decommissioning Plan dpm Disintegration per minute DQO Data Quality Objectives EH&S Environment, Health and Safety ENTOMB Entombment option ER Environmental Report FEMA Federal Emergency Management Agency FGR Federal Guidance Reports FSS Final Status Survey HASP Health and Safety Program HAZWOPER Hazardous Waste Operations and Emergency Response v

BMRC Decommissioning Plan Revision 0 HEPA High Efficiency Particulate Air HIC High Integrity Container HMR Hazardous Materials Regulation HP Health Physics HRA High Radiation Area HSA Historical Site Assessment HVAC Heating, Ventilation, and Air Conditioning ICS Incident Command System IF Irradiation Facility INEEL Idaho National Engineering and Environmental Laboratory ISO International Standards Organization LLRW Low-Level Radioactive Waste LSA Low Specific Activity LSC Liquid Scintillation Counter m meter MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MDC Minimum Detectable Concentration MW t Megawatt Thermal mrem Millirem MSHA Mine Safety and Health Administration MTR materials-testing-reactor N-16 Nitrogen-16 NaI Sodium Iodide NEI Nuclear Energy Institute NEPA National Environmental Policy Act NIOSH National Institute for Occupational Safety and Health NIST National Institute of Science and Technology NMSS Nuclear Material and Safety and Safeguards NRC United States Nuclear Regulatory Commission NSTC Nuclear Science and Technology Center N2S2 Nevada National Security Site NUREG Nuclear Regulatory Guide NYS DOH New York State Department of Health pCi/g picocuries per gram P.E. Professional Engineer vi

BMRC Decommissioning Plan Revision 0 PM Project Manager PuBe Plutonium-Beryllium PULSTAR Pulse Training Assembled Reactor OSHA Occupational Safety and Health Act QA Quality Assurance QAPP Quality Assurance Program Plan QC Quality Control RA Restricted Area RDSC Reactor Decommissioning Safety Committee RP Radiation Protection RPM Radiation Protection Manager RSO Radiation Safety Officer RWP Radiation Work Permits SAFSTOR No-Action Alternative SNM Special Nuclear Material SCR Site Characterization Report SSC Structures, Systems, and Components SUNY State University of New York TEDE Total Effective Dose Equivalent TLD Thermoluminescent dosimeter UB University at Buffalo UBPM University at Buffalo Project Manager WRS Wilcoxon Rank-Sum yr year vii

BMRC Decommissioning Plan Revision 0 1.0

SUMMARY

OF DECOMMISSIONING PLAN 1.1 Introduction The Buffalo Materials Research Center (BMRC) is owned by The State University of New York (SUNY) at Buffalo (UB) and is located on the southern edge of the South Campus of UB off of Rotary Drive in Buffalo, Erie County, New York. The BMRC was a Research and Test Reactor Facility with a pool-type reactor that was designed and constructed by AMF Atomics between 1959 and 1961. The initial criticality date for the reactor was March 24, 1961, while the last day of operation occurred on June 23, 1994. The unit has been in Possession Only status since June 6, 1997. In 1998 the unused fuel was shipped to North Carolina State University. In 2005 the spent fuel was shipped to the Idaho National Engineering and Environmental Laboratory. The BMRC is being decommissioned because there is no future need for the facility.

UB is about 20 miles south of the Canadian border at Niagara Falls, New York, 90 miles west of Rochester, New York and 80 miles from the southern border of New York and Pennsylvania. UB is about five miles east of the Niagara River which is a border with Canada. Figure 1-1 shows the location of the BMRC on the UB South Campus.

Figure 1-1 Location of BMRC on the UB South Campus 1

BMRC Decommissioning Plan Revision 0 Decommissioning Plan Synopsis This Decommissioning Plan (DP) has been prepared in accordance with Chapter 17 of the Nuclear Regulatory Commission (NRC) NUREG-1537 Part 1, Guidelines for Preparing and Reviewing Applications for Licensing of Non-Power Reactors and NUREG-1757, Consolidated Decommissioning Guidance. This Decommissioning Plan provides guidance on the general process and methods to be used to safely decontaminate, remove and dispose of radioactive materials, equipment, systems, components, and soil associated with the BMRC. Decommissioning activities will result in the complete removal of the BMRC structures from the site location allowing an unrestricted release of the BMRC site by the NRC and be granted termination of the NRC license as allowed by Title 10, Code of Federal Regulations (CFR), Part 20, Section 1401 (10 CFR 20.1401).

Decommissioning will be performed in a safe and orderly manner to prevent undue radiation exposure to the workers, public, and environment. Remediation and demolition tasks include 1) the removal of the reactor core and activated bioshield; 2) removal of reactor and facility systems containing residual radioactive materials; 3) removal of ventilation systems; 4) remediation of concrete surfaces; and 5) disposition of building structural concrete and steel. All structures of the building and facility piping are planned to be removed in their entirety.

A final status survey (FSS) will be implemented to demonstrate compliance with default derived concentration guideline levels (DCGLs) in support of the unrestricted release of the site and termination of the facility.

1.2 Background Gener al The reactor at the BMRC was operated by SUNY at Buffalo under an operating license from the NRC.

The reactor at UB was placed into operation in 1961 under License Number R-77 and operated until 1963 with materials-testing-reactor (MTR) fuel elements with a maximum steady-state power level of 1 megawatts thermal (MW t ). In 1964 the reactor was shut down and the core and control systems were modified so that the reactor could operate with Pulse Training Assembled Reactor (PULSTAR) fuel at power levels of up to 2 MW t .

The UB ceased operations of the reactor on June 23, 1994, and shipped the unused PULSTAR fuel to North Carolina State University in 1998. In 2005, UB with the assistance of the U.S. Department of 2

BMRC Decommissioning Plan Revision 0 Energy (DOE) shipped the used PULSTAR fuel to the DOE at the Idaho National Engineering and Environmental Laboratory (INEEL).

The BMRC currently maintains two NRC Licenses, a Possession Only License (R-77) and a Special Nuclear Material License (SNM-273) for possession of a Plutonium-Beryllium (PuBe) neutron source; however, this source has been transferred to the UBs New York State Department of Health (NYS DOH) agreement state license. In addition to the PuBe source, the NYS DOH license (#1051) authorizes the possession of byproduct materials for calibration and check sources as well as byproduct materials from reactor operations.

Histor y The reactor was constructed in 1959 and 1960 under Atomic Energy Commission (AEC) Construction Permit No. CPRR-39 and went critical on March 24, 1961. The original licensed power was 1 MW t with an original core loading of MTR fuel. The MTR Fuel was replaced with PULSTAR fuel in 1964 and the facility license was amended to operate at 2 MW t .

During its operating history, the BMRC was used for training and education, transient fuel performance testing, nuclear component testing and calibration, materials radiation damage research, isotope production, and neutron interrogation through activation analysis, radiography and delayed fission assay.

The following is a chronological history of significant licensing events:

1959 - Initial construction commenced under AEC License # R-77.

Licensee name: The Western New York Nuclear Research Center 1961 - Construction completed in January 1961 - Initial criticality achieved on March 24 at 2051 hours0.0237 days <br />0.57 hours <br />0.00339 weeks <br />7.804055e-4 months <br /> Full power first achieved on June 1 at 1220 hours0.0141 days <br />0.339 hours <br />0.00202 weeks <br />4.6421e-4 months <br /> 1964 - Reactor modifications to utilize PULSTAR fuel; Used MTR fuel transferred to McMaster University in Ontario, Canada in April Initial criticality of PULSTAR core achieved June 22 at 2112 hours0.0244 days <br />0.587 hours <br />0.00349 weeks <br />8.03616e-4 months <br />; 1973 - Facility name change: Nuclear Science and Technology Center (NSTC) 1984 - Vertical 10k tank placed into service; Tank Farm placed into wet layup 1985 - Facility name change: Buffalo Materials Research Center 1991 - Reactor and Primary Coolant System modified to present configuration 1994 - Reactor activities suspended; Last shutdown on June 23 at 0600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br />; 1997 - NRC license R-77 amended to Possession Only Status 1998 - Unused PULSTAR fuel shipped to North Carolina State University in July 2005 - Used fuel shipped to Idaho National Engineering and Environmental Laboratory 3

BMRC Decommissioning Plan Revision 0 Reactor Descr iption The reactor is a PULSTAR heterogeneous open-pool type water cooled reactor that used solid 6%

enriched uranium dioxide fuel clad in Zircaloy similar to commercial nuclear power reactor fuel. The core was cooled by forced convection at higher power levels and by natural convection at lower power levels.

The coolant is purified water and the reflector is water and/or graphite.

The core is immersed in a 13,000-gallon aluminum alloy lined pool surrounded by both high density and regular density concrete. Original design specifications used water to cool the core. The water exited the reactor at the pool bottom and was then pumped to a 5,000-gallon Nitrogen-16 (N-16) holdup tank for decay. The water then cycled through a heat exchanger connected to an external cooling tower located on the southeast side of the Administration Building (Laboratory Wing). The cooled water was then circulated back to the top of the pool.

In October 1977 a leak developed in the primary cooling pipe. The leak was repaired by June of 1978 and reactor operations resumed. In 1989 a reoccurrence of a leak at the point where the coolant piping penetrated the bottom (horizontal portion) of the reactor liner directly under the reactor led to the decision that modifications needed to be made. These modifications were completed in 1991 and included adding a second aluminum alloy liner on the Neutron Deck level and rerouting the coolant piping through the original twelve-inch beam tube, just above the floor of Neutron Deck level of the Containment Building.

The coolant piping on the Neutron Deck is shielded by high density concrete blocks until it penetrates through the containment wall into the N16 Tank Vault.

All nuclear fuel has been removed from the site. The reactor and its associated components remain in the pool in the normal operational configuration. Reactor components removed during repairs completed in 1991 were stored behind concrete block shielding on the Neutron Deck. These reactor components were removed during the pre-decommissioning cleanout and buried at the EnergySolutions Clive, UT facility..

Associated with the reactor within the Containment Building are the following Irradiation Facilities:

x Flux Trap x Isotope Tube x Vertical Standpipe x Thermal Column x Pneumatic Conveyor x Electron Accelerator 4

BMRC Decommissioning Plan Revision 0 BMRC Facility The BMRC Facility consists of the reactor described above, the Containment Building which encloses the reactor and associated facilities related to use of the reactor, as well as the Administrative Building which contains offices, class rooms, and laboratories. The Administration Building is also referred to as the Laboratory Wing.

Containment Building The reactor is housed in a tri-level Containment Building as shown in an artists rendering of the interior of the original Containment Building in Figure 1-2. The three levels (from top to bottom) of the Containment Building include the Control Deck, the Gamma Deck, and the Neutron Deck. Plan views of the Control Deck, Gamma Deck, and Neutron Deck are shown on Figure 1-3, Figure 1-4, and Figure 1-5 respectively. The Containment Building was constructed of reinforced concrete and in a right cylinder shape. The cylindrical Containment Building is approximately 75 feet in diameter and 52 feet high with walls that are two-foot thick reinforced concrete. The roof of the Containment Building varies from four to nine inches in thickness and is supported by concrete beams. The walls and the foundation of the Containment Building are constructed on bedrock.

Within the Containment Building are the following additional facilities:

x Hot Cell and Hot Cell Work Room (Room 105) x Dry Chamber (Neutron Deck Level) x Charpy Cell (Neutron Deck level) (Removed) x Hot and Warm Chemistry Labs (Rooms 103, 104, and 108) x Neutron Activation Analysis and Counting Lab (Room 109) x Other Labs (Rooms 202, and 203) x Electronics Shop (Room 200) 5

BMRC Decommissioning Plan Revision 0 Figure 1-2 Artists Rendering - Interior Cut Away View of the Reactor and Containment Figure 1-2 Legend:

1. Bioshield 21. Hot Cell

2. Reactor Plenum 22. Hot Cell Crane and Rail

3. Thermal Column Plug 23. Pass Tube Isolation Valve

4. Beam Tube 24. Hot Cell Lead Shutter

5. Thermal Column Support 25. Hot Cell Lead Shutter Drive Mechanism

6. Thermal Column Graphite Blocks 26. Hot Cell Remote Manipulators

7. Thermal Column Support Plate 27. Radiation Shielding Window

8. Thermal Column 28. Hot Cell Instrumentation Access Port

9. Flux Trap 29. Hot Cell Removable Access Plug

10. Beam Tube Chamber 30. Dry Chamber

11. Reactor 31. NA

12. Coolant Outlet Plenum 32. Removable Metal Plate Floor Opening

13. Coolant Outlet Piping Cover

14. Reactor Pool Lining 33. Pass Tube Isolation Valve Hand Wheel

15. Vertical Standpipe 34. Control Deck Airlock (Access to/from

16. Pneumatic Conveyor Tube Administration Building)

17. Isotope Tube

18. Isotope Tube Drive Mechanism

19. Operating Platform

20. NA 6

BMRC Decommissioning Plan Revision 0 Control Deck The top level of the Containment Building is called the Control Deck. A plan view of the Control Deck is shown on Figure 1-3. The Control Deck houses a dual-hook overhead crane (with a 10-ton and 2-ton hook) at the ceiling level. Crane access to the Gamma Deck is through three floor openings shown on Figure 1-3. Two of the floor openings are covered with removable metal plates. In addition to the reactor pool, the Control Deck houses the Reactor Control Room (Room 205), Superintendents Office (Room 207), two labs (Rooms 202 and 203), Fan Room (Room 204), another office (Room 206), Electronics Shop (Room 200), and the New Fuel Storage Room in the airlock (Room 201B). The airlock provides for access to the upper level of the Administration/Laboratory Building.

Figure 1-3 Control Deck Plan View 7

BMRC Decommissioning Plan Revision 0 Gamma Deck The middle level of the Containment Building is called the Gamma Deck. A plan view of the Gamma Deck is shown on Figure 1-4. The Hot Cell, Hot Cell Work Room (Room 105), Locker Room (Room 106), Room 107, Hot Chemical Lab (Room 108), Medium Chemical Lab (Room 104), Activation Analysis Lab (Room 109), Lower Fuel Vault Room (Room 101A), and Chemical Lab (Room 103) are located on the Gamma Deck. This level also has an Airlock (Room 101) for access to the bottom level of the Administration Building and a truck door for access to the asphalt drive located on the southern side of the Containment Building.

The Truck Door provides vehicular and equipment access to the outside. Prior to 1979, the Truck Door area housed a waste compaction system. Wastes generated at the BMRC (NSTC at the time) and under the site-wide radiological program were collected and compacted for volume reduction in this system.

Analysis activities involving the use of mercury in conjunction with radiological constituents occurred in the Activation Analysis room. Mercury may be present in the fume hoods in this room and in the low points or traps in waste drain lines.

Figure 1-4 Gamma Deck Plan View 8

BMRC Decommissioning Plan Revision 0 Neutron Deck The lower level of the Containment Building is called the Neutron Deck. A plan view of the Neutron Deck is shown on Figure 1-5. The Neutron Deck is an open area divided by fencing into two sections.

The enclosed (fenced) portion of the Neutron Deck, shown in red hatch in Figure 1-5, is currently a high radiation area (HRA). The HRA is being used to store reactor components removed from the reactor during repairs made in 1991 due to a leak in the reactor liner. These reactor components (Thermal Column Nosepiece, Dry Chamber Nosespiece, Grid Plate Bolts, and Neutron Beam Tube ends) were replaced with new items of similar. The rerouted Primary Coolant piping, 18,000 lbs of Depleted Uranium Shield block, and access to the beam storage tubes is also in the HRA. The beam storage tubes extend ten feet through the Containment Building wall into the surrounding soil. They are used as storage-for-decay of high dose reactor components or experiments. At the point where the tubes exit the Containment Building wall, they are approximately ten feet underground.

Figure 1-5 Neutron Deck Plan View 9

BMRC Decommissioning Plan Revision 0 Reactor , Bioshield, Dr y Chamber, and Hot Cell The reactor, bioshield, dry chamber, and hot cell are located inside the Containment Building and portions are located on the various levels of the building. Figure 1-6, Figure 1-7, and Figure 1-8 contain plan views of these items on each level of the Control Deck, Gamma Deck, and Neutron Deck respectively. Figures 1-9 and 1-10 provide elevation views of the reactor and bioshield. These figures reflect the original setup of the reactor and do not include modifications to the reactor tank and coolant piping in 1977 or the tank modification completed in 1991.

The bioshield consists of high density concrete with metal aggregate in lieu of typical stone aggregate.

The bioshield section on the Gamma deck facing the Truck Door consists of removable concrete blocks set in place by the facility crane. The remaining sections of the bioshield were cast in place after the entire reactor tank, beam tubes, and piping were constructed. It should be noted that floors and floor supports were poured in place concurrently with the Containment Building walls. Therefore, the Gamma Deck and Control Deck floors do not appear to be structurally supported by the reactor bioshield. The structural independence between the bioshield and the floors should be verified by a New York State Professional Engineer using the original facility design drawings and the chronological historical photos before bioshield demolition.

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BMRC Decommissioning Plan Revision 0 Figure 1-6 Reactor Plan View - Control Deck Level 11

BMRC Decommissioning Plan Revision 0 Figure 1-7 Reactor Plan View - Gamma Deck Level 12

BMRC Decommissioning Plan Revision 0 Figure 1-8 Reactor Plan View - Neutron Deck Level Figure 1-9 Reactor/Bioshield Elevation View (Northeast Direction) 13

BMRC Decommissioning Plan Revision 0 14

BMRC Decommissioning Plan Revision 0 Figure 1-10 Reactor/Bioshield Elevation View (Southeast Direction) 15

BMRC Decommissioning Plan Revision 0 Ventilation System (Containment Building)

Air from the Heating, Ventilation, and Air Conditioning (HVAC) system from the portions of the Containment Building which housed the medical facilities, hot cell, hot chemistry laboratory, beam storage tubes, rabbit tubes, and the thermal column was discharged via underground piping to the 167-foot high stack on the Mackay Heating Plant located southeast of the BMRC facility (see Figure 1-1).

Portions of this piping have been removed during renovation of Rotary Road as detailed in the Historical Site Assessment (HSA).

The remainder of the Containment Building ventilation passes through a bank of roughing and absolute filters prior to being discharged from a point on the southeast side of the Containment Building roof. Per the facility Technical Specifications, ventilation ducts were designed with hydraulic dampers that would isolate the Containment Building from the Administration Building and the outside areas if alarms indicated an airborne event.

Liquid Waste System (Containment Building)

Prior to 1984, liquid waste from the hot cell drain, hot chemistry laboratory, and service area drains was collected in the two 250-gallon tanks located in the tank farm (see Figure 1-5). The tank farm is an underground facility located southeast of the Containment Building. Liquid waste from the Gamma Deck and Control Deck laboratories, change area sinks, and Truck Lock door drain were collected in the two 600-gallon tanks located in the tank farm.

Administration Building (Laboratory Wing)

The Administration Building may also be referred to as the Laboratory Wing in this document. Within the Administration Building are the following facilities:

x Vertical Accelerator (Room 115A) x Horizontal Accelerator and Counting Cave (Room 118) x Health Physics Office and Counting Room (Room 114) x Low Level Labs with Fume Hoods (Rooms 213, and 215) x Machine Shop (Room 113)

Upper Level - Administration Building A plan view of the upper level of the Administration Building is shown on Figure 1-3. The upper level of the Administration Building housed the Mens Locker and Change Room (Room 209) and Mens Bathroom (Room 209B), Womens Locker and Change Room (Room 221) and Womens Bathroom 16

BMRC Decommissioning Plan Revision 0 (Room 221A), Conference Room (Room 212), offices (Room 208, Room 210, Room 214, and Room 216), Reception Area (Room 208A), Janitors Closet (Room 209A) and two low level laboratories (Room 213 and Room 215). There is a fume hood in Room 213 and Room 215. Based on characterization, there has been little to no radiological impact to the upper level of the Administration Building.

Basement Level - Administration Building A plan view of the basement level of the Administration Building is shown in Figure 1-4. On the basement level of the Administration Building is housed a Horizontal Accelerator (Room 118) with an adjacent cave area. This level also contained a laboratory (Room 115), Classroom (Room 117), Health Physicist Office and Counting Room (Room 114), The Vertical Accelerator Room (Room 115A),

Machine Shop (Room 113), and Air Conditioning Equipment Room (Room 111). The Electrical Service Room (Room 110) is also located on this level. A utility tunnel is located north of the Electrical Service Room (Room 110). The utility tunnel connects to the Campus Service Tunnel system and is locked to prevent access.

Subbasement - Administration Building A plan view of the subbasement of the Administration Building is shown on Figure 1-5. The subbasement level is located under the western portion of the basement. The eastern portion of the basement level is supported by caissons and grade beams. The subbasement contains the 5,000-gallon Nitrogen-16 Holdup Tank Room (Room N16), Cooling Water Equipment Room and Sump Pit Room (Room N03), lower Air Conditioning Equipment Room (Room N02) and adjacent area way, Air Handling Duct Room (Room N04), and access to the vertical 10,000-gallon tank room (Room N05) as well as the vertical 10,000-gallon tank room. The floor of Room N16 is located on bedrock.

Ventilation System (Administration Building)

Air from the HVAC system located in the Administration Building is vented through high efficiency particulate air (HEPA) filters to the roof of the building except for the Machine Shop (Room 113) and Vertical Accelerator Room (Room 115A) fume hoods. Air from these two fume hoods is vented via a duct on the southeast side of the Administration Building.

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BMRC Decommissioning Plan Revision 0 Liquid Waste System (Administration Building)

Liquids from the bathrooms, floor drains, sinks, and fume hood drains in the Administration Building are discharged to the South Campus sanitary system. Experiments in the Administrative Building have involved use of H-3 and C-14 in the upper level laboratories (Room 213 and Room 215) fume hoods and H-3 in the Health Physicist Office and Counting Room (Room 114) fume hood.

1.2.1 Reactor Decommissioning Overview UB plans to completely dismantle the reactor and its ancillary support systems, remove radioactive materials from the BMRC, and demolish the facility prior to requesting the termination of license R-77.

The facility is being demolished because there is no future need for the facility.

Most of the reactor components and systems are either activated or contaminated and will need to be segregated from non-radiological components and surfaces so that they can be disposed of as low-level radioactive waste (LLRW). The BMRC is primarily constructed of concrete and a vast majority of the waste generated during the decommissioning will be concrete rubble from demolition activities. Waste is disposed of based on characterization for radiological constituents. Based on characterization results, it is anticipated that almost all of the structural concrete will be sent for concrete recycling or will be sent to a local industrial landfill. Some concrete, primarily the floor slabs of the subbasement and the Neutron Deck are anticipated to be disposed of as Bulk Survey for Release (BSFR) waste due to the potential for volumetric contamination. Notable exceptions are the concrete associated with the reactor bioshield on the Neutron Deck level as well as the reactor liner and internal components. Some decontamination is expected to be required for the reactor components, bioshield, and hot cell concrete structures that were out of the main neutron activation zone so that the majority of concrete and metal from these structures will meet the requirements of BSFR waste. The decommissioning cost estimate assumes that the reactor liner and internal components are to be disposed of as Class A LLRW as shown in Appendix A, BMRC Activations Analysis and Component Characterization, developed by WMG, Inc. of Peekskill, NY while the old Control Blades are to be disposed of as Class B mixed radwaste.

The following major decommissioning tasks are necessary for site release. The sequence in which these tasks occur may vary (see Section 2.3 for more details):

x Pre-Decommissioning Cleanout - removal of loose legacy equipment and materials (Completed) x Isolation and removal of inactive systems (Some prior to approval of DP by the NRC) x Removal of hazardous materials (lead, cadmium) and abatement of asbestos containing material (ACM) (Bulk of the materials to be removed prior to approval of DP by the NRC) x Perform supplementary characterization (prior to approval of DP by the NRC) 18

BMRC Decommissioning Plan Revision 0 x Installation of temporary systems and preparation of the facility for decommissioning operations (potentially prior to approval of DP by the NRC) x Removal of the irradiated reactor components x Removal of and disposition of water in the reactor x Removal of and disposition of the reactor liner x Decontamination of concrete associated with reactor, bioshield, and hot and dry cell x Removal of the reactor, bioshield, and hot and dry cell x Segregation, packaging and shipment of materials according to radioactivity levels x Removal of auxiliary systems (rabbit system, water purification, ventilation) x Decontamination of building surfaces x Interior release survey using Reg Guide 1.86 limits x Demolition of the BMRC facility x Perform the FSS on BMRC footprint x Submit required reports that demonstrates to the NRC that the facility meets the release criteria x Request license R-77 termination x Restore site: Backfill void created by removing below grade structures for future use by UB.

The FSS is to be developed by UB using the criteria provided in NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) (NRC 2000). An overview of the FSS is provided in Section 4.0 with the Final Status Survey Plan to be submitted and approved prior to commencement.

1.2.2 Estimated Cost The decommissioning cost estimate is summarized in Table 1-1. It is estimated that approximately $8.9 million is required to decommission the BMRC. The cost estimate includes the Demolition Contractor (DC) costs, subcontractor costs, and other direct costs. The cost includes the costs of shipping and disposing of waste and the final status survey. The cost estimate is subject to cost increases because there are unforeseeable elements of cost within the defined project scope. A contingency of 20% is included in the decommissioning costs in Table 1-1 to ensure that sufficient funds are available to cover costs that may result from unanticipated conditions or unforeseeable elements in the project scope. Typically, these include factors such as waste disposal rates or increased waste volumes from undiscovered or uncharacterized areas. In addition, the time duration between the development of the DP and the inception of decommissioning activities can influence the costs associated with changes in the economy and regulatory requirements.

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BMRC Decommissioning Plan Revision 0 Table 1-1 Decommissioning Cost Estimate Item Description TOTAL No.

1 Project Planning $ 238,440 2 Mobilization and Training $ 67,620 3 Facility Preparation $ 235,480 4 Neutron Deck remediation and disposal $ 1,792,092 5 Gamma Deck remediation and disposal $ 221,882 6 Control Deck remediation and disposal $ 141,946 7 Administration Building remediation and disposal $ 51,038 8 Waste and HVAC Systems remediation and disposal $ 983,429 9 Ancillary systems removal and disposal costs $ 406,000 10 Impacted Soil excavation and disposal costs $ 99,270 11 Waste Packaging and Transportation $ 849,800 12 Demolition of Administration and Containment Buildings $ 1,801,700 13 Final Status Survey and Report $ 248,980 14 Backfill and Site Restoration $ 223,995 15 Demobilization $ 61,620 SUBTOTAL $7,423,292 Contingency @ 20% $ 1,484,658 TOTAL DECOMMISSIONING COST $ 8,927,950 1.2.3 Availability of Funds In accordance with 10 CFR 50.75 (e)(1)(iv), UB is a state institution and as such will provide financial assurance with a statement of intent containing a cost estimate for decommissioning, indicating that funds will be obtained when necessary. The decommissioning of the BMRC is fully funded in the current UB Capital Budget that runs through June 2013.

1.2.4 Program Quality Assurance 1.2.4.1 Independent Contracting and Oversight The UB will select a qualified DC to assist in the physical remediation and dismantlement of the BMRC reactor and associated facilities. In addition, the UB has contracted with a Design and Oversight Contractor (DOC) to provide the UB with Decommissioning Project Management, Health Physics, Safety, Quality, and other oversight support as needed to augment the UB staff in order to meets its license and regulatory obligations during decommissioning. The selected DOC is Enercon Services, Inc.

(ENERCON). The selection of the DOC is discussed in Section 2.6.

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BMRC Decommissioning Plan Revision 0 The dual-contractor approach has been selected by the UB to assure that decommissioning work activities are subject to continuous and independent compliance oversight. The DOC will report directly to the UB Project Manager (UBPM) for project schedule and cost while reporting to the UB/BMRC Environment, Health and Safety Department in matters affecting quality or safety. The DOC will be responsible for developing a Quality Assurance Project Plan (QAPP) appropriate for the decommissioning of the BMRC and the final status survey. The QAPP will be reviewed and approved by the UB.

The DOC shall monitor UB contracted work to assure compliance with the UBs reactor license, the approved DP, and applicable contract requirements and regulations. The DOC will also provide site-specific training [developed and approved by the BMRC Director and Radiation Safety Officer (RSO)] to the selected contractors to assure their familiarity with the requirements of the UBs reactor license and the BMRC DP.

1.2.4.2 Established Minimum Qualifications for DC The UBs selection of qualified contractors for the DC role will consider the following criteria as a minimum:

a) Ability of the firm to perform the required tasks as demonstrated by past nuclear reactor demolition experience (5 years) of the firm, its field supervisors, and its teaming partners or subcontractors as appropriate.

b) Qualifications of key individuals based on education, training, experience and professional licensure.

c) The compliance record of the contractor, teaming partners, and key subcontractors; including records of any violations or fines for non-compliance with federal, state or local regulations.

d) Safety record of the contractor and key subcontractors to include a review of past projects and Occupational Safety and Health Act (OSHA) reportable incidents.

e) Relevant experience of contractor, teaming partners, and key subcontractors, particularly with regard to research reactor decommissioning.

f) References and recommendations from previous customers/completed projects.

g) Financial stability of the contractor and key subcontractors to complete the project and their ability to meet the minimum insurance and bonding requirements.

h) Evaluation of the DCs proposed Technical Approach for site demolition.

1.2.4.3 Quality Assurance Program Plan (QAPP)

The DOC will be responsible for preparing and implementing a QAPP to control the decontamination and decommissioning of the BMRC. The QAPP will incorporate industry and regulatory requirements 21

BMRC Decommissioning Plan Revision 0 applicable to decommissioning project planning and management, decontamination, dismantling, demolition, radiological sampling, analysis, and surveys. The QAPP shall describe the organization, document approval, and quality assurance (QA) aspects for the BMRC decommissioning project, including the following minimum requirements:

a) A clear description of the project organization, including a definition of the authorities and responsibilities of key personnel involved in the decommissioning.

b) QA actions to be implemented during the BMRC decommissioning project.

c) A list of implementing policies and procedures the DOC and DC will use to complete the project, including detailed project work plans for critical project evolutions.

d) Work approval, communication, and authority guidelines, including guidelines for the interface of the DC, the DOC, and the BMRC/UB.

e) Document control for UB approved work plans and procedures.

f) A plan for performing self-monitoring and independent audits of the DOC and DCs activities.

g) Maintenance of records for project work, calibration, and characterization data.

1.2.4.4 Shipping The DOC will identify appropriate treatment and disposal facilities for wastes generated from the decommissioning project. The DOC will develop a procedure for compliant shipping of BMRC wastes, identifying methods of compliance with 10 CFR 71 subpart H and Department of Transportation (DOT)

Hazardous Materials Regulations (HMR) as appropriate. The DOC is responsible for waste sizing, characterization, segregation, and packaging. The QAPP and associated procedures will address quality requirements for waste preparation and packaging activities.

1.2.4.5 Final Status Survey QAPP The DOC will develop a QAPP specific to the performance of the FSS and associated documentation (e.g.

characterization information used in the design of the final survey). The FSS QAPP will be reviewed and approved by the UB. The FSS QAPP will incorporate the appropriate regulatory requirements applicable to the planning and conduct of radiological surveys necessary for the termination of the BMRC license and the release of the site for unrestricted use. The DOC will perform the FSS in accordance with the QAPP and related procedures, and will submit the completed FSS to the UB for approval and submittal to the NRC.

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BMRC Decommissioning Plan Revision 0 2.0 DECOMMISSIONING ACTIVITIES The objective of the BMRC decommissioning activities is to remove licensed radioactive materials from the facility and surrounding grounds necessary to obtain NRC approval for release to unrestricted use of the property and be granted termination of the NRC license. The decommissioning pathway described in this plan is intended to meet the necessary requirements to achieve this objective.

2.1 Decommissioning Alternatives There are four alternatives available to, and considered by, the UB: 1) the No-Action alternative (SAFSTOR); 2) the entombment option (ENTOMB); 3) complete decontamination and structure demolition (DECON-A); and 4) complete decontamination and release of the structure (DECON-B). The DECON options are recommended by the NRC for non-power reactors. The selected alternative is DECON-A. The four alternatives are summarized in the following sections.

2.1.1 SAFSTOR Alternative The SAFSTOR alternative would involve the same potential risks and environmental impacts as the proposed action, but for a much greater time period. The alternative would require that the UB maintain current radiological controls, site security, required licensing, a reactor administrator, and the utilities until the DECON option was implemented. The UB would also have to incur all of the expenses associated with maintaining the facility. This alternative would require that the UB apply for and obtain an extension to the current NRC operating license. The SAFSTOR alternative would leave the site as unusable space that would delay any beneficial reuse; and the potential for environmental contamination would still exist. This alternative is not environmentally preferable.

2.1.2 ENTOMB Alternative The ENTOMB alternative places the same requirements on the UB as the SAFSTOR alternative with the additional requirement that the radioactive material be placed in a long-lived structure, i.e., concrete, or that the current structure proven adequate for long-term storage. This alternative delays the impacts of the proposed action and the potential for environmental contamination would continue to exist. This alternative is not environmentally preferable.

2.1.3 DECON-A Alternative The facility records and current facility characterization reveal minimal facility contamination from past reactor operations. This alternative requires the site to be released and restored for unrestricted use. The reactor, the containment building, and the administrative building under this option will be disassembled 23

BMRC Decommissioning Plan Revision 0 and the radioactive material removed to meet the release criteria. The BMRC facility is located in the center of a growing university and the land area could be reused for future construction; therefore, complete decontamination and demolition (D&D) of the BMRC facility, DECON-A, is the preferred option. This alternative poses minimal risk and impacts to the environment as described the Decommissioning Environmental Report (ER).

2.1.4 DECON-B Alternative The DECON-B alternative is similar to DECON-A, however, the structure would not be removed after remediation. The risk and impacts to the environment are the same as DECON-A for this alternative, but the UB has determined that the reuse of the building is not in its best interests. Additionally, a final status survey of the soils and bed rock under a standing structure is more complicated and requires a significantly greater level of surveying and sampling than releasing an open excavation. This alternative is not preferred by the UB.

2.2 Facility Radiological Status 2.2.1 Facility Operating History The BMRC reactor operated from March 24, 1961 until June 23, 1994. The last of the spent fuel assemblies were removed from the facility in 2005. During the operating lifetime of the BMRC it served several functions related to education, health services, neutron activation analysis, radiological chemistry, radiological biology, nuclear science technology information dissemination, services to industry, and contractual research and design. A sampling of some of the tasks performed at the BMRC is listed below.

x Testing of nuclear industry components in a radiological environment.

x Study radiation effects on the solid state properties of carbon.

x Study radiocarbon dating techniques.

x Production of superfluids (He6).

x Investigation of Mercury contamination in fetal brain tissue.

x Study of Mercury in adult brain tissue.

x Cadmium toxicity studies.

x Samples tested for gun powder residue to support a criminal justice case.

x Samples tested for environmental pollution in air, water, and soil.

x Radiation sensitization of microorganisms.

x Isotope production for medical applications.

x The use of radioactive isotopes for visualization of bone lesions for medical applications.

x Testing of neutron detectors and other radiation detection equipment for industry application.

x Neutron activation analysis of blood serum.

x Neutron activation analysis for testing lead contamination and for forensic testing.

x Calibration of instruments for health physics applications.

x Demonstration of safe disposal of liquids, solids, and gaseous radioactive waste.

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BMRC Decommissioning Plan Revision 0 2.2.2 Current Radiological Status of the BMRC Facility The BMRC facility has been impacted from the operation of the research reactor. However, the radiological characterization performed to prepare for the development of this DP indicates that practices employed to minimize the spread of radioactive material were effective and therefore the impacts have been limited.

A majority of the waste generated from the demolition of the BMRC facility is expeced to be eligible for release. It should be possible to recycle and reuse the concrete as aggregate or dispose of the concrete as construction and demolition (C&D) waste in a local permitted landfill. These options reduce both the overall cost of decommissioning and the associated safety risk because: (1) the risk of a transportation accident is significantly reduced with local recycling versus interstate transportation for BSFR or Class A LLRW disposal; (2) reduced cost of transportation to a local disposal facility versus distant LLRW landfills or processing facilities; (3) if the concrete is disposed of as C&D waste, the disposal rates are significantly lower than for BSFR waste or Class A LLRW.

Appendix B, BMRC Site Characterization Report, provides a detailed status of the radiological conditions at the BMRC. Characterization efforts were guided by a thorough HSA and a characterization plan.

Characterization activities included subsurface soil samples down to bedrock, fixed location alpha and beta measurements on building surfaces, waste tank samples, loose activity smears, smears for hard to detect radionuclides, bioshield concrete samples, scan measurements on building surfaces, 10 CFR Part 61 distribution samples, and a gamma walkover survey. Analytical results from subsurface soil samples adjacent the reactor tank were less than the laboratory minimum detectable concentrations (MDCs) for contaminants of concern. In addition, indications of residual activity were not found outside of areas described in the HSA that were expected to contain elevated levels of radioactive materials.

The information obtained from both the site characterization effort and the WMG Activation Analysis and Component Characterization provides sufficient data to demonstrate that it is unlikely that significant quantities of residual radioactivity have gone undetected in the BMRC. Table 2-1 below provides an estimate of the types and quantities of radioactive materials.

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BMRC Decommissioning Plan Revision 0 Table 2-1 Estimated Volumes and Masses of Radiologically Impacted Components and Systems Waste Volume(a)

Component/System Primary Material type (ft3) Mass (lb)

Reactor Internals/Components Steel, aluminum, graphite A 200 5000 (b)

Reactor and Bioshield Walls Concrete A 2,700 401,600 Reactor Foundation Concrete A 1,300 185,400 Hot Cell Concrete A 1,700 244,500 (b)

Reactor Tank liner 6061-T6 alloy (aluminum) A 100 12,500 Old Control Blades Nickel Plated Permalloy B (mixed) 2 200 Containments Walls Concrete C&D 19,000 2,714,200 Foundation (Neutron Deck) Concrete BSFR 1,600 232,800 Neutron Deck Ceiling Concrete C&D 5,100 757,200 Neutron Deck Concrete BSFR 2,200 328,700 Gamma Deck Ceiling Concrete BSFR 7,200 1,188,000 Bioshield Walls >12 Ft Concrete C&D 2,700 324,000 Hot Cell Concrete BSFR 22,600 3,389,900 Gamma Deck Rooms Concrete C&D 5,500 812,700 Gamma Deck Room 109 Non-concrete C&D 2,400 403,300 Control Deck Ceiling Concrete C&D 1,400 204,800 Control Deck Rooms Concrete C&D 300 6,800 Laboratory Wing Concrete C&D 6,000 896,400 Laboratory Wing Non-concrete C&D 1,600 234,200 Ancillary Structures Concrete C&D 31,700 4,603,800 HVAC System Steel BSFR 6,800 168,800 Soil Soil BSFR 6,600 982,100 Total A 6,000 849,000 Total B (mixed) 2 200 Total BSFR 47,000 6,290,300 Total C&D 75,700 10,957,400 (a) Estimated packaged volume.

(b) Assumes a maximum of 12 ft. height in the activation region is removed.

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BMRC Decommissioning Plan Revision 0 2.2.3 Release Criteria The decommissioning alternative selected by the UB is DECON-A. This section provides the specific radiological criteria that will be applicable for unrestricted release of the site and termination of NRC license R-77.

DECON-A includes the removal of activated and contaminated materials, equipment and components.

Because the BMRC facility will be demolished, the release criteria for the FSS will be for the soil and bedrock that remain on site. The remaining soil and rock surfaces will be released to less than the NRC required 25 millirem (mrem) annual Total Effective Dose Equivalent (TEDE) following guidance contained in MARSSIM. The release criterion will be determined to have been met by demonstrating surface or volumetric activities meet their respective NRC screening values as presented in Tables 2-2 and 2-3.

2.2.3.1 Release Criteria for Structures, Systems and Components Structures, Systems, and Components (SSCs) destined for reuse, recycling, or disposal as clean waste will be shown to be free of detectable surface contamination in accordance with the guidelines provided by the NRC in IE Circular 81-07 (NRC 1981). Monitoring for residual radioactivity will use instrumentation and techniques (background radiation levels, scan speed, counting times) necessary to detect activity no greater than 5,000 dpm/100cm2 total and 1,000 dpm/100cm2 removable beta/gamma contamination. All instruments shall be calibrated with radiation sources having an energy spectrum and instrument response consistent with the radionuclides being investigated. If alpha contamination is suspected, appropriate residual radioactivity measurements capable of detecting alpha activity no greater than 100 dpm/100cm2 fixed and 20 dpm/100cm2 removable will be used. Per IE Circular 81-07, the potential dose impacts to members of the public is less than 5 mrem/yr to the maximally exposed individual which will meet the NRC release criteria of 25 mrem/yr. This release criteria is also described in NRC Regulatory Guide (Reg Guide) 1.86, Termination of Operating Licenses for Nuclear Reactors, and is known throughout the industry as the Reg Guide 1.86 release criteria.

Properly calibrated survey instrumentation with known efficiencies capable of measuring the radionuclide of concern will be used for release surveys. Removable contamination wipes may be measured in a liquid scintillation counter (LSC) or a wipe/filter counter such as the Ludlum Model No. 3030E or equivalent.

For surface tritium contamination, only removable contamination will be assessed because of the difficulties in measuring total tritium surface contamination directly (ISO 1988). If a removable fraction of 10% is assumed (ISO 1988), analysis for removable tritium must have a minimum detection limit no 27

BMRC Decommissioning Plan Revision 0 greater than 500 dpm/100cm2 so that the total (fixed plus removable) required detection limit of 5,000 dpm/100cm2 is not exceeded. Tritium wipes shall be measured in an LSC.

2.2.3.2 Release Criteria for Soils Characterization data indicates that there has been minimal impact to subsurface soils from operation of the reactor. Soils remaining in place will be shown to meet an annual TEDE of 25 mrem/yr by demonstrating the average volumetric activity is less than the NRC screening values as found in Appendix B of NUREG-1757, Consolidated Decommissioning Guidance and presented in Table 2-2.

The screening values are listed in picocuries per gram (pCi/g) of material.

Table 2-2 NRC License Termination Screening Levels for Surface Soils NRC Screening Value Radionuclide (pCi/g)

Ag-108m 8.2*

Am-241 2.1 C-14 12 Co-60 3.8 Cs-137 11 Eu-152 8.7 Eu-154 8 H-3 110 Ni-63 2,100 Pu-238 2.5 Pu-239/240 2.3 Ra-226 0.7 Sr-90 1.7 See Section 2.2.3.4 for derivation of Ag-108m screening value.

2.2.3.3 Release Criteria for Bedrock Surfaces During decommissioning, the entire BMRC structure will be demolished and removed from the site. The bottom concrete slab of the Containment building, i.e. the Neutron Deck floor, was poured directly on the bedrock. After the building is demolished, only the bedrock will remain. The bedrock is not volumetrically contaminated; therefore, the FSS will employ survey methods utilized for building 28

BMRC Decommissioning Plan Revision 0 surfaces using the NRC screening values for building surfaces as found in Appendix B of NUREG-1757, Consolidated Decommissioning Guidance and presented in Table 2-3. The bedrock will be prepped prior to the FSS by removing loose materials, i.e. concrete dust, pieces, and dirt. Release criteria for radionuclides not in Table 2-3 will be consistent with the Reg Guide 1.86 values stated in Section 2.2.3.1.

The amount of loose radioactive materials shall not exceed 10% of the release criteria.

Table 2-3 NRC License Termination Screening Levels for Building (Bedrock) Surfaces NRC Screening Value Radionuclide (dpm/100cm2)

Ag-108m 17,000*

C-14 3,700,000 Co-60 7,100 Cs-137 28,000 H-3 120,000,000 Ni-63 1,800,000 Sr-90 8,700 See Section 2.2.3.4 for derivation of Ag-108m screening value.

2.2.3.4 Derivation of Ag-108m Screening Criteria The screening criterion provided by the NRC does not list a value for Ag-108m. The NRCs source of calculating a screening value is the computer code DandD developed by Argonne National Laboratory.

This computer code does not have Ag-108m listed in its radionuclide library. In order to avoid the lengthy and time burdensome task of calculating site-specific DCGLs using probabilistic RESRAD runs, the UB DOC has calculated a Ag-108m screening value based on the screening value calculation of Ag-110. The Ford Nuclear Reactor at the University of Michigan experienced the same technical issue with Ag-108m and was the first to calculate screening values for Ag-108m.

Isotopes of an element behave the same chemically; therefore, Ag-108m will behave the same as Ag-110 in the environment and human body as well as have the same exposure pathways. The only difference between the two radionuclides in the calculation of their screening value in the DandD computer code is the Dose Conversion Factor (DCF) for each radionuclide.

The DCFs for Ag-108m and Ag-110m were gathered from Federal Guidance Reports (FGR) #11 and #12.

The ratio of the two DCFs was used as a dose factor to multiply to the dose calculated for Ag-110m in the 29

BMRC Decommissioning Plan Revision 0 DandD model for each individual exposure pathways. This resulted in a dose per pCi/g for Ag-108m for each exposure pathway. The values are listed in Table 2-4 below.

Table 2-4 Ag-108m Surface Soil Screening Value Ag-110m Dose Dose factors Ag-108m Dose Pathway (mrem/1.0 pCi/g) (Ag-108m/Ag-110m) (mrem/1.0 pCi/g)

External Exposure 4.58E+00 0.581 2.66E+00 Inhalation 1.22E-05 0.493 6.01E-06 Sec. Ingestion 9.62E-05 0.705 6.78E-05 Agricultural 5.59E-01 0.705 3.94E-01 Drinking Water 1.36E-20 0.705 9.59E-21 Irrigation 1.34E-19 0.705 9.45E-20 Surface Water 7.26E-22 0.705 5.08E-22 Cumulative mrem/(pCi/g) 3.05 (pCi/g)/mrem 0.327 Screening Value @ 25 mrem/y 8.19 pCi/g Using the Ag-108m/Ag-110m Dose Factor for each exposure pathway, the resultant cumulative dose per unit of volume (mrem per pCi/g) is 3.05 mrem. This means that an average concentration of 1 pCi/g will result in a dose to the public of 3.05 mrem per year. The inverse of this number provides the average concentration that will result in a dose of 1.0 mrem/yr which is 0.327 pCi/g. The average concentration values normalized to 25mrem/yr is 8.19 pCi/g.

In the same process described above, the Ag-108m screening value for building surfaces was calculated as shown in Table 2-5 below.

Table 2-5 Ag-108m Building Surface Screening Value Ag-110m Dose Dose factors Ag-108m Dose Pathway 2 (mrem/1000 dpm/100 cm ) (Ag-108m/Ag-110m) (mrem/1000 dpm/100 cm2)

External Exposure 2.34E+00 0.604 1.41E+00 Inhalation 1.05E-01 0.493 5.18E-02 Ingestion 7.88E-03 0.705 5.56E-03 Cumulative 1.47E+00 Screening Value @ 25 mrem/y 17,000 dpm/100 cm2) 30

BMRC Decommissioning Plan Revision 0 2.3 Decommissioning Tasks 2.3.1 Activities and Tasks 2.3.1.1 Preparation Tasks Prior to approval of the DP by the NRC, several activities are scheduled to be conducted to prepare the reactor facility for decommissioning. Preparation of the facility and the removal of non-reactor structures, systems, and components can advance the overall decommissioning schedule. The following sections describe the preparation tasks scheduled to be performed by the UB, the DOC, and other contractors.

2.3.1.1.1 Pre-Decommissioning Cleanout Miscellaneous loose materials and equipment that were located in the BMRC were packaged for disposal to prepare for the decontamination and demolition phase of decommissioning. An inventory of the loose materials and equipment was developed and a Request for Proposal was developed. The items sent for disposal were items that did not require any type of dismantlement, i.e., unbolting from fixed surfaces or torch cutting. The miscellaneous materials included, but not limited to, shield block, lead bricks, lead aprons, old tank samples, inoperable equipment, reactor components replaced during the 1989 rebuild. By removing the miscellaneous equipment and materials from the BMRC, additional work and storage space is available for the D&D phase of decommissioning and additional characterization of the areas that are currently occupied by the waste can be completed.

All materials were sent to a licensed radioactive materials processor for proper disposition of the material, (i.e., BSFR, Class A Radioactive Waste Disposal, lead recycling) as authorized by the processors radioactive materials license.

The waste is scheduled to be disposed of prior to the submittal of this DP to the NRC and the UB intent to perform this activity was discussed with the NRC and NYS DOH prior to commencement. Project records, shipping manifests, and inventories are available for inspections as necessary.

2.3.1.1.2 Isolation and Removal of Inactive Systems Inactive systems that are not required by either technical specifications, or for safety, or for support of decommissioning activities may be isolated, de-energized, and drained prior to approval of the decommissioning plan. This stage is often referred to as interference removal. These systems will be removed from the facility to avoid cross contamination during the removal of activated or impacted items.

Removal of inactive systems is necessary because typically these systems interfere with the physical dismantling and demolition activities associated with decommissioning. Any inactive radioactively 31

BMRC Decommissioning Plan Revision 0 impacted systems will be removed at the direction of the DOC. The remaining (required) systems will be removed after approval of the DP by the NRC.

2.3.1.1.3 Removal of Hazardous Waste and ACM The UB may contract with a qualified licensed contractor who will be responsible for the disposal of hazardous waste and ACM in compliance with federal, state of New York, and local regulations and requirements. A licensed asbestos abatement contractor will remove ACM utilizing industry practices.

The DOC will provide oversight and radiation protection support as necessary during the removal of ACM. To date, characterization activities have not identified radioactively contaminated ACM, however, radiation protection (RP) support will be provided to ensure proper RP controls are in place should they be needed. Similarly, for hazardous waste, the DOC will provide the workers with the appropriate radiation safety training required commensurate with the potential for exposure to radioactive materials and RP support to ensure that radioactively impacted hazardous material will be controlled and segregated from hazardous material that is not radioactively impacted.

2.3.1.1.4 Facility Preparation for Decommissioning Temporary systems needed to support decommissioning activities such as temporary power, portable lighting, temporary ventilation systems, and air monitoring systems may be installed prior to approval of the DP. If facility modifications are required to facilitate waste handling, they will be performed following approval by the UB.

2.3.1.2 Tasks and Activities for BMRC Decommissioning The DOC will modify and/or augment existing BMRC and UB procedures to accommodate the decommissioning effort. Additionally, best available industry techniques will be employed during the D&D of the reactor SSCs. These techniques may include, but are not limited to, the use of long handled and standard tools, hydraulic cutters, torches, plasma arc torches, wire saws, needle guns, jackhammers, hand-scabblers, high pressure and ultra high-pressure sprayers and cutters. Any cost-effective tool or technique that achieves the goal of D&D while maintaining the principles of as low as reasonably achievable (ALARA) and considers the end state of the waste will be employed. Procedures and work packages will be developed by the DC designating the specific tool or technique to be employed. These procedures will be reviewed by the DOC and approved by the UB. Methods that will minimize the spread of contamination such as portable high efficiency particulate air (HEPA) ventilation systems or an encapsulation medium may be used. The utilization of temporary shielding will be used to maintain 32

BMRC Decommissioning Plan Revision 0 personnel exposures ALARA. Water will be left in the reactor pool/liner during the removal of the reactor internal components.

It is anticipated that decommissioning will initially focus on the items that have the greatest potential to be radioactively impacted. This includes the reactor and its components, the bioshield, the dry cell, and the hot cell. Then, it is anticipated that any radioactively impacted systems will be remediated. After these items have been remediated, a majority of radioactively impacted material will have been removed from the facility. Therefore, decommissioning can proceed at a more efficient pace because the potential for cross-contamination and undue exposure will be limited due to the fact that very little radioactive material will remain. This approach will also allow work to be performed ALARA.

The following activities are presented as an overview (of decommissioning) and the UB may elect to perform the activities in a different sequence. ALARA, safety, cost, and scheduling may dictate that a different sequence be used by the UB. Current BMRC staff has performed the first three activities previously during the rebuild completed in 1991. Their experience and documentation from the rebuild will be used to perform the activities in a manner that is safe and ALARA.

2.3.1.2.1 Removal of the Operating Platform The operating platform is located above the reactor. It will be removed to allow better access to the reactor and its components. The operating platform will be disposed of based on the release criteria in Section 2.2.3 and Section 2.2.5.

2.3.1.2.2 Removal of Reactor Components Reactor core components include, but are not limited to, the following: the core frame, plenum, core blades and shrouds, thermal column frame, and sleeve for the fission chamber. Reactor components, irradiation facility (IF) tubes for experiments and other items located within approximately three feet of the core for extended periods of time have been activated to levels that will require disposal as LLRW.

The upper portion of reactor components, i.e. Control Blade drives, IF tubes, and instrumentation housings, may be surveyed and released or disposed as BSFR waste.

Removal of the reactor components should be performed while water is still in the reactor pool to provide shielding so that the work can be performed while keeping exposure ALARA. Long-handle tools and remotely operated equipment may be used to disassemble the components. The components will be lifted using the overhead crane or other device, and will be placed in LLRW containers, with appropriate shielding to keep dose rates ALARA and within DOT and disposal facility acceptance limits.

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BMRC Decommissioning Plan Revision 0 2.3.1.2.3 Water Removal from the Reactor Tank The water in the reactor tank will be drained, filtered, and discharged to the sanitary system, as approved by the UB using the existing liquid waste system. Radioactive sediments on the bottom of the tank can be removed by agitating the sediments to remove them during pumping, for subsequent filtering.

Alternatively, the sediments may be left still and removed manually after the tank is drained. The tank water will be filtered to remove the suspended solids, stored, and sampled according to current procedures to allow discharge of the water to the sanitary system. Any water from any ancillary systems, such as the water purification system or water in the tanks in the Tank Farm, should also be processed in the same manner. Filter media and water not meeting the discharge limits will be managed as LLRW.

2.3.1.2.4 Removal of Activated Portions of the Reactor Tank Liner In the rebuild completed in 1991, a liner was installed in the bottom half (Neutron Deck Level) of the reactor tank. Both the tank and the liner are made of an aluminum alloy (6061-T6). The aluminum liner may be removed prior to destructive dismantling of the bioshield. It has been assumed that the inner and outer liner will require disposal as LLRW. However, UB may elect to survey the newer liner material and only dispose of material as LLRW only in the activation region. Since the bioshield was poured around the completed reactor tank, it is not feasible to remove the original tank separately from the bioshield.

2.3.1.2.5 Removal of Reactor Bio-Shield, Dry Cell, and Hot Cell After the activated core components (and possibly the reactor liner) have been removed the concrete associated with the reactor bioshield, dry cell, and hot cell will be removed by destructive methods, i.e.

cutting and/or hydraulic shear. As described in Section 1.2, the bioshield, the dry cell, and the walls/ceiling of the hot cell were poured independently from the Containment Building floors and walls.

The concrete waste from the demolition of the portions of these structures between the Neutron Deck and the bottom of the Gamma Deck will be disposed of as Class A LLRW. The remaining concrete waste from these structures is expected to be BSFR waste.

The destructive removal technique for the bioshield will be cutting by diamond wire, or equivalent, into pieces that can be lifted by the BMRC overhead crane having a maximum capacity of 10 tons. The dry cell and hot cell walls will primarily be removed by diamond wire cutting, but portions may be cut by a rotary saw or blunt force demolition.

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BMRC Decommissioning Plan Revision 0 Cut points will be verified by a New York Licensed professional engineer (P.E.) to ensure the structural integrity of the BMRC is not compromised. Concrete dust from destructive removal operations will be controlled and continuously monitored for industrial hygiene and radiological concerns.

2.3.1.2.6 Removal of HVAC system and other impacted systems HVAC system removal may be necessary prior to Containment Building demolition. Radioactively impacted systems may be disposed of as LLRW if decontamination is not possible or cost effective.

Minimally impacted systems will be disposed of as BSFR waste. Any systems that were maintained to support decommissioning activities will be de-energized at this stage to prepare for demolition of the Containment Building. The storm water and sanitary systems will be removed up to the junction point where they tie into other UB South Campus systems. The tanks used in the current and former liquid waste management systems will be disposed of as LLRW. The soil and gravel covering the tanks, along with the concrete enclosure and berms are expected to be disposed of as BSFR waste.

2.3.1.2.7 Waste Systems Removal The waste systems consist mainly of the Tank Farm which lies in the subsurface adjacent to the containment building and extends towards the southwest of the containment building (see Figure 1-5) and the 10K Vertical Waste Holding Tank. A smaller 1,000-gallon holding tank in the subbasement that is in line with the 10K Vertical tank will also be removed. The removal of the waste systems will include the removal of tanks and the piping associated with both waste systems. The tanks will be size reduced by industry standard methods and disposed of as LLRW. Due to the expected residual radioactive material in the tanks, the tanks will be flushed prior to removal and local ventilation will be used unless the tanks are shipped offsite for processing at a radioactive waste processing facility.

2.3.1.2.8 Demolition of Containment Building The Containment Building will be safely demolished after the work described in Sections 2.3.1.2.1 through 2.3.1.2.7 are complete, the activated concrete is removed, and building surfaces are remediated where necessary to meet the limits specified in Section 2.2.3.2. The demolition debris, i.e., concrete and rebar, will be sent to a local recycler or to a local C&D landfill. An exception to this is the bottom slab, i.e., the Neutron Deck floor, which is expected to be disposed of by BSFR due to impacts from the reactor leaks. Although cores through the Neutron Deck floor showed no indication of contamination, it is deemed too costly to perform the amount of surveying on the bottom of the concrete surface required for release to a recycler.

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BMRC Decommissioning Plan Revision 0 Industry standard demolition techniques will be used to demolish the BMRC. The concrete dust will be continuously monitored and controlled to applicable federal, state, and local limits. Air monitoring will also be conducted periodically for radiological parameters.

2.3.1.2.9 Demolition of Administration Building The Administration Building will be demolished in conjunction with the Containment Building because there is no future need for the BMRC. It is anticipated that some minor decontamination will be required prior to demolition. It is also anticipated that if there are any radioactively impacted systems they will be removed prior to demolition. The building is mostly composed of concrete construction materials so a majority of the waste will be concrete rubble. The demolition debris will be disposed of in the same manner as the Containment building, i.e., the subbasement slab as BSFR and the remaining debris to a local recycling company or local C&D landfill.

2.3.1.2.10 Remediation of Soils and Bedrock Characterization has not shown that there are radioactively impacted soils at the BMRC. The samples collected in the Room N16 were scrapings of dirt on the bedrock. It is possible that a small amount of soil around the waste tanks have elevated levels of radioactive materials. This material is to be remediated and characterized for proper disposal. In addition, removal of impacted bedrock is required in Room N16 due to historical releases. The amount of bedrock to be remediated is expected to be approximately one cubic yard with residual material to be removed by scabbling or chiseling of the bedrock.

2.3.2 Schedule The project duration, from DP approval to completion of the FSS, is approximately 15 months. The proposed decommissioning schedule is presented in Figure 2.1. Changes to the schedule may be made at the UBs discretion to respond to resource management, availability of a radioactive waste burial site, interference with ongoing UB activities, ALARA considerations, and/or temporary on-site radioactive waste storage operations. The current project schedule is consistent with other recently completed university research reactor decommissioning projects.

The schedule includes activities from project planning through NRC review of the Final Status Survey Report and license termination. It is based on a 5-day work week with the number of work days listed for each activity.

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BMRC Decommissioning Plan Revision 0 Figure 2-1 Proposed Project Schedule 2.4 Decommissioning Organization and Responsibilities The decommissioning of the BMRC facility is under the technical supervision of the UBs Environment, Health and Safety (EH&S) Department who is responsible for assuring that decommissioning activities are conducted in a safe manner and within the requirements of the NRC License, this DP, the UBs Radiation Protection Program, and other state and federal laws and regulations. Personnel in the UB EH&S Department that are providing supervision of the decommissioning for licensing, radiation safety and industrial safety operated the BMRC before it was shutdown in the positions of Director, Operations Manager, Reactor Engineer, and RSO. The UB has appointed a project manager to oversee the decommissioning process.

The Organizational Chart in Figure 2-2 at the end of this section shows the lines of reporting at the UB BMRC. The Decommissioning Radiation Safety Staff reports to the RSO for radiation safety and licensing functions which are represented by the dashed reporting line. Additionally, the Decommissioning Industrial Safety Staff will report to the UB Reactor Engineer for safety and compliance reporting for industrial safety concerns.

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BMRC Decommissioning Plan Revision 0 2.4.1 University Project Manager The following duties, at a minimum but not limited to, will be assigned to the UBPM.

x Selecting a Demolition Contractor in accordance with UB procurement guidelines.

x Overseeing the DOC and DC performance relative to the terms of their contract.

x Overseeing the DOC and DC performance relative to plans and procedures.

Minimum required qualifications and experience requirements for the UBPM are as follows:

x Bachelor's degree in Architecture, Civil, Electrical, Mechanical or Structural Engineering or related field AND five (5) years of construction experience which included one year of construction supervisory experience; OR, x Nine years of progressively responsible construction experience which included one year of construction supervisory experience; OR, x Any equivalent combination of experience, training and/or education approved by the UB Human Resources department.

2.4.2 BMRC Director The BMRC Director has management responsibility and technical oversight for facility operations and radiation safety programs. The following duties and responsibilities are assigned to the BMRC Director, or designee:

x Ensuring that decommissioning activities are performed in compliance with applicable regulations and license conditions.

x Approval of plans and procedures required for decommissioning x Reviewing and submitting to the Reactor Decommissioning Safety Committee (RDSC) needed changes and subsequent plans and procedures that do not change the original intent of the DP or result in an unreviewed safety question.

x Communicating with the NRC, New York State Agencies, and UB Administration.

Minimum required qualifications and experience requirements for the BMRC Director are as follows:

x Advanced degree (MS or PhD) in Nuclear Engineering or related discipline or equivalent experience and five (5) years experience in nuclear reactor operations and/or decommissioning.

x Familiarity with NRC License R-77, the Decommissioning Plan, the BMRC Radiation Protection Program, and with applicable federal and state regulations.

x Trained at the level required by the UB RP Program to be in possession of radioactive materials of the types known to be present at the licensed reactor site.

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BMRC Decommissioning Plan Revision 0 2.4.3 Radiation Safety Officer The UB RSO is responsible for monitoring and overseeing radiological safety at the reactor and BMRC.

The RSO has the responsibility and authority to stop any plan or activity that has the potential to result in an unacceptable radiological hazard. The following duties and responsibilities will be assigned to the UB RSO, or designee:

x Supervise the implementation of the UB radiation safety program by DOC radiation safety staff.

x Review and approve radiation safety procedures.

x Review and approve radiation work permits.

Minimum required qualifications and experience requirements for the RSO are as follows:

x Advanced degrees (MS or PhD) and/or certified in Health Physics with 10 years experience in radiation safety and health physics.

x Familiarity with NRC License R-77, NYS Radioactive Materials Licenses, the Decommissioning Plan, the BMRC Radiation Protection Program, and with applicable federal and state regulations.

2.4.4 Reactor Decommissioning Safety Committee The function, responsibilities, and makeup of the RDSC are defined in the Technical Specifications (UB Research Reactor license R-77). Among those responsibilities are, but not limited to:

x Approval of plans required for decommissioning.

x Review and approval of proposed changes to the facility, procedures and Technical Specifications and DP.

x Determination of whether a proposed change to the DP would constitute an unreviewed safety question or a change in the Technical Specifications as required by 10 CFR 50.59, and review and approval of required safety analysis.

2.4.5 Design and Oversight Contractor The DOC is responsible for the direct field management of the BMRC decommissioning and assisting the UB in the administration of the industrial and radiological safety programs. The duties assigned to the DOC PM include, but are not limited to, the following:

x Manage the safe and regulatory compliant implementation of the UB DP and FSS.

x Assist the UBPM in overseeing the DC performance relative to subsequent plans and procedures.

x Development, implementation, and associated costs of a Radiation Safety program compliant with 10 CFR 20.

x Development, implementation, and associated costs of an OSHA compliant Health and Safety Program.

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BMRC Decommissioning Plan Revision 0 x Review and approval of work plans and procedures necessary for the safe and compliant decommissioning of the BMRC.

x Assist the UB in acquiring applicable permits for radiological waste disposal and transportation.

The DOC will assign a PM with the following minimum required qualifications and experience:

x B.S. Degree and 10 years combined experience in decommissioning, project management, and radiation safety with at least five (5) of the years specific to nuclear reactor decommissioning.

x Experience in the decommissioning of university research and test reactors.

x Familiarity with applicable federal and state regulations, MARSSIM and NRC decommissioning guidance.

See Section 2.6 for additional criteria used in the selection of the DOC.

2.4.6 Demolition Contractor The DC is responsible for the development, implementation, and associated costs of work plans and procedures for the safe and compliant demolition of the UB BMRC. The DC will report directly to the DOC Project Manager for implementation of the DP, but will be contracted directly to the UB. The DC is also assigned the following responsibilities:

x Development and safe implementation of work plans and procedures for the demolition of the BMRC.

x Compliance with the BMRC DP.

x Compliance with the BMRC Decommissioning Radiation Safety program.

x Compliance with the BMRC Decommissioning Health and Safety Program.

The minimum required qualifications and experience for the selection of a DC is discussed in Section 2.6.

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BMRC Decommissioning Plan Revision 0 Figure 2-2. Organizational Chart Vice President for University Life and Services Associate VP for University Facilities Reactor Decommissioning Director, Safety Committee Director Facilities, Environment, Health Planning and Design and Safety Services FP&D Manager of BMRC Director Support Asst. Director EH&S Reactor Engineer BMRC/UB University Project Manger BMRC Operations Radiation Safety Manager Officer BMRC Site UB Radiation Decommissioning Representative Project Manager Safety Staff (DOC)

Site Decommissioning Decommissioning Radiation Protection Superintendent Safety Manager Manager (DC) (DOC)

(DOC)

Decommissioning Decommissioning Decommissioning Decommissioning Craft Personnel Work Planner(s) Safety Engineer(s) HP Staff (DC) (DC) (DOC) (DOC) 41

BMRC Decommissioning Plan Revision 0 2.5 Training Programs 2.5.1 General Site Training A general training program will be designed and implemented by the DOC and approved by UB to provide orientation to project personnel and meet the requirements of 10 CFR 19, Notices, Instructions, and Reports to Workers: Inspection and Investigations. General site training will be required for personnel assigned on a regular basis to the D&D project. General site training will include but is not limited to:

x Project orientation, security, and access control - such as BMRC Operating Procedure # 56; BMRC Unescorted Access Training for Non-BMRC Personnel x Introduction to radiation protection x Quality assurance x Industrial safety x Emergency procedures x Packaging and transport of radioactive materials The following are examples of additional training that may be required:

x Radiation Worker Training will meet the requirements identified in the BMRC Decommissioning Radiation Protection Plan.

x Hazardous Waste Operations and Emergency Response (HAZWOPER) training will be required for personnel engaged in hazardous substance removal or other activities that potentially expose them to hazardous substances and health hazards.

x Respirator Training and Fit Testing will be performed according to the BMRC Decommissioning Respiratory Protection Program.

x Hazard Communication Training will be provided to personnel exposed to hazardous or potentially hazardous materials.

x Hearing Conservation Training will be provided on the effects of noise on hearing and the purpose, advantages, disadvantages, and attenuation of various types of hearing protective devices.

x Permit-Required Confined Space Entry Training will be required for personnel entering confined spaces.

x Lockout/Tagout Hazardous Energy Control Training for hazardous energy control.

x Trenching and Excavation Training for the purpose of determining the safety and stability of excavations.

For specific tasks that require state licensing or other special qualifications, the qualifications will be reviewed by the DOC PM or Safety Manager. If additional radiation safety training is required, it will be provided by the site RSO.

2.5.2 Radiation Worker Training The reactor D&D operations will be managed by the DOC and performed by the DC. As such, the DOC will be responsible for the radiation worker training of BMRC decommissioning contractors and 42

BMRC Decommissioning Plan Revision 0 subcontractors in accordance with the requirements of 10 CFR 19 with the final approval of qualifications by the RSO. The Decommissioning Radiation Protection Manager (RPM) will be responsible for on-site radiation safety training of workers and verifying pervious training and qualification as approved by the RSO. The DOC's radiation safety training program will be administered by the RPM who will approve, along with the RSO, training materials and qualification of workers. The RSO will provide dose monitoring badges such as thermoluminescent dosimeters (TLD).

The minimum radiation safety training provided to any worker will include, but is not limited to the following subjects:

x Principles of radiation protection x Radiation monitoring techniques x Radiation monitoring instrumentation x Emergency procedures x Radiation hazards and controls x Concepts of radiation and contamination x Provisions of 10 CFR 19 and 20 x NRC license conditions and limitations x Reporting requirements for workers x Biological effects of radiation x Radiation control zone procedures x Radiation Work Permits (RWP)

A written exam will be required to demonstrate proficiency with the radiation worker training topics.

Radiation worker training will also include a practical factors demonstration and evaluation. This evaluation will include a review of the following:

x Proper procedures for donning and removing protective clothing and equipment.

x The ability of the worker to read and interpret self-reading and/or electronic dosimeters (if used).

x Proper procedures for entering and exiting a controlled area, including proper frisking techniques Persons who have documented equivalent radiation worker training from another site or employer within the previous twelve (12) months may be waived from taking the training but must take the written and practical factors examinations. Radiation Worker training will be refreshed on an annual basis or at the discretion of the RSO.

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BMRC Decommissioning Plan Revision 0 2.6 Contractor Assistance The UB has already selected a qualified DOC for the design and oversight of the BMRC decommissioning based on a combination of the following items:

1. Response and demonstration of prior experience (five years minimum) on the tasks listed in the project scope.

i. Project Management ii. Pre-Decommissioning Clean Out iii. Site Characterization iv. Decommissioning Plan Development

v. NRC and New York State Regulatory Interface vi. Interference Removal vii. Decommissioning Drawings and Specifications viii. Decommissioning Oversight Activities ix. Final Status Survey planning and implementation

2. Prior University Research and Test Reactor Decommissioning Experience

3. Experience of Personnel Dedicated to the Project

4. State Agency/NRC Experience

5. Design / Technical Experience

6. Minority Status / Staffing In addition to the DOC, the UB will select a qualified contractor to perform the physical demolition of the BMRC. In selecting the contractor, the UB will produce a request for proposal, which will define the qualifications and experience necessary for prospective DCs and subcontractors. Prior history and performance of the prospective contractor on non-power reactor or similar decommissioning projects will be used to help the UB select a qualified contractor to perform the facility D&D.

Qualification and experience requirements that will be used to evaluate the DC are to have at least 5 years prior experience in radiological site decontamination. Specific experience in the decontamination and demolition of test reactors, power reactors, and/or materials licensed sites will be required by the prime contractor. DC submittals of project descriptions, references, and other supporting information will be required prior to contract award. Specific DC project management documentation will be required in the areas of work plan development, training, QA, work management, reactor dismantlement and decontamination, waste packaging, waste shipping, work documentation, and supporting the preparation of the final decommissioning project report. The minimum expectation of the UB for a DC is verification of company experience in these tasks, proof of financial viability, and bonding capacity.

The UB will require the selected DC to support the project with an experienced Superintendent, supervisors, and Work Planner(s) with at least five years prior experience in radiological site decontamination and demolition with preference in test reactors in a college campus setting. Specific individual experience will be required in the areas of work plan development, training, QA, work 44

BMRC Decommissioning Plan Revision 0 management, reactor dismantlement and decontamination, waste packaging, waste shipping, work documentation, and supporting the preparation of the final decommissioning project report. The minimum expectation of the UB for a DC is verification of personnel experience in these tasks plus a commitment to provide experienced personnel for the duration of the project.

2.7 D&D Documents and Guides This decommissioning plan was prepared using the guidance and format specified in Chapter 17 of NUREG-1537 (NRC 1996). The radiological criteria for license termination to allow unrestricted use will be as set forth in 10 CFR 20, Subpart E. The decommissioning project will also be administered according to the applicable section of the following regulations and regulatory guidance documents:

Code of Feder al Regulations 10 CFR 19 Notices, Instructions and Reports to Workers; Inspections 10 CFR 20 Standards for Protection Against Radiation 10 CFR 30 Rules of General Applicability to Domestic Licensing of Byproduct Material 10 CFR 50 Domestic Licensing of Production and Utilization Facilities 10 CFR 51 Licensing and Regulatory Policy and Procedures for Environmental Protection 10 CFR 71 Packaging of Radioactive Materials for Transport and Transportation of Radioactive Materials Under Certain Conditions 29 CFR 1910 Occupational Safety and Health Standards 29 CFR 1926 Occupation Safety and Health Standards for Construction 49 CFR 170-199 Department of Transportation Hazardous Materials Regulations NRC Regulator y Guides 1.86 Termination of Operating Licenses for Nuclear Reactors 1.187 Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments 8.2 Guide for Administrative Practices in Radiation Monitoring 8.7 Occupational Radiation Exposure Records Systems 8.9 Acceptable Concepts, Models, Equations and Assumptions for a Bioassay Program 8.10 Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Reasonably Achievable 8.13 Instruction Concerning Prenatal Radiation Exposure 8.15 Acceptable Programs for Respiratory Protection 45

BMRC Decommissioning Plan Revision 0 NUREG-Ser ies Publications 1505 A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys 1507 Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions 1549 Using Decision Methods for Dose Assessment to Comply with Radiological Criteria for License Termination, Draft 1575 Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) 1640 Radiological Assessments for Clearance of Materials From Nuclear Facilities 1756 Technology, Safety, and Cost of Decommissioning Reference Nuclear Research and Test Reactors 1757 Consolidated Decommissioning Guidance Additional project-specific documents will be developed by the DOC and/or the UB prior to starting the D&D project. Such documents may include:

x RP and ALARA Plan x Site Health and Safety Plan x QAPP x Waste Management Plan x FSS Plan x Specific Task Plans 46

BMRC Decommissioning Plan Revision 0 3.0 PROTECTION OF WORKERS AND THE PUBLIC 3.1 Radiation Protection The RPM, under the direct supervision of the RSO, will administer the Health Physics Program (also referred to as the Radiation Protection Program) specifically developed for the BMRC Decommissioning as described in Section 3.1.2. The DOC will supplement the radiation protection program with detailed plans and procedures specific to the radionuclides listed in Table 3-1. The RSO, the RPM and Decommissioning Health Physics (HP) Staff will be responsible for implementing ALARA principles; providing radiation worker training; establishing administrative-level occupational and public dose limits; monitoring personnel for occupational exposures; controlling exposures; providing and maintaining radiation monitoring equipment; performing radiation surveys and monitoring; and maintaining records and generating reports as necessary to comply with regulatory and licensing requirements.

Table 3-1 BMRC Radionuclides of Concern as determined in the Site Characterization Report Radionuclide Half Life (yr) Emission Area(s) of Interest Ag-108m 4.18E+02  Soil; Tank Water; SSCs Am-241 4.32E+02  Tank sediment C-14 5.73E+03 Laboratory areas Co-60 5.27E+00  Soil; SSCs; Bioshield Cs-137 3.01E+01 * Soil; SSCs; Bioshield Eu-152 1.36E+01  Soil; SSCs; Bioshield Eu-154 8.59E+00  Soil; SSCs; Bioshield H-3 1.23E+01 Soil; Bioshield and Tank Water Ni-63 1.00E+02 Soil; SSCs; Bioshield Pu-238 8.78E+01  Tank sediment Pu-239 2.41E+04  Tank sediment Pu-240 6.60E+03  Tank sediment Ra-226 1.60E+03  Discrete sources Sr-90 2.88E+01 SSCs; Ventilation systems; Soil

HPLVVLRQIURP%D-137m progeny 47

BMRC Decommissioning Plan Revision 0 3.1.1 Ensuring ALARA Radiation Exposures The DOC will prepare an RP and ALARA Plan that will incorporate provisions for minimizing occupational and public radiation exposures. This Plan will describe specific administrative and engineering controls that will be put in place during specific D&D project activities. Examples of administrative and engineering controls include limiting access to certain areas, mock-up training, use of remote-handling devices, temporary shielding, containment structures, portable HEPA filtered ventilation, and specialized protective equipment and respiratory protection.

3.1.2 Health Physics Program The project Health Physics Program will be implemented under the authority of the RSO with the assistance of the DOC RPM. The Health Physics Program will satisfy the following commitments that should be established by the Radiation Protection Program:

x Implement the procedures defined in the Radiation Protection and ALARA Plan.

x Ensure radiological safety of the public, occupationally-exposed personnel, and the environment.

x Monitor radiation levels and radioactive materials.

x Control the distribution and release of radioactive materials.

x Maintain potential exposures to the public and occupational radiation exposure to individual within administrative limits and the regulatory limits of 10 CFR 20 and ALARA.

x Monitor personnel internal and external exposure in accordance with 10 CFR 20 requirements.

3.1.2.1 Project Health Physics Program - General UB has procedures in place that will be implemented during the BMRC Decommissioning Project.

Additional Health Physics procedures may be required and will be developed and approved in accordance with UB policy and procedure.

UB senior management is readily accessible to ensure timely resolution of difficulties that may need to be addressed prior to regularly scheduled meetings. The RSO, while organizationally independent of the project staff, is a direct report to the BMRC Director. He also has full authority to act in all aspects of protection of workers and the public from the effects of radiation. Conduct of the BMRC Decommissioning Project HP program will be evaluated according to UB policy.

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BMRC Decommissioning Plan Revision 0 3.1.2.2 Audits, Inspections, and Management Review During Decommissioning Project work, aspects of the Project may be assessed and reported by the DOC's Quality Assurance Department, through audits, assessments and inspections of various aspects of decommissioning performance, including HP, as described in Section 1.2.4.3 Program Quality Assurance.

Audits of the BMRC Health Physics program are conducted in accordance with the requirements of 10 CFR 20. These audits will include aspects of the BMRC Decommissioning Project. Additional assessments or management reviews may be performed when deemed appropriate by the EH&S Director.

HP equipment and instrumentation suitable to permit ready detection and quantification of radiological hazards to workers and the public will be chosen to ensure the validity of measurements taken during remediation and formal release surveys. The selection of equipment and instrumentation will be based upon detailed knowledge of the radiological contaminants, concentrations, chemical forms and chemical behaviors that are expected to exist as demonstrated during radiological characterization, and as known from process knowledge of the working history of the BMRC. Equipment and instrumentation selection will also take into account the working conditions, contamination levels and source terms that are reasonably expected to be encountered during the performance of decommissioning work, as presented in this Plan.

The following sections present details of the equipment and instrumentation planned for use during the decommissioning. It is anticipated that through retirement of worn or damaged equipment/

instrumentation or increase in quantities of available components or instruments, that new technology will permit upgrades or, at minimum, like-for-like replacements. UB is committed to maintaining conformance to minimum performance capabilities stated in this Plan whenever new components or instruments are selected.

3.1.2.3 Criteria for Selecting Equipment and Instrumentation for Conduct of Radiation and Contamination Surveys and Personnel Monitoring A sufficient inventory and variety of instrumentation will be maintained on site to facilitate effective measurement of radiological conditions and control of worker exposure consistent with ALARA, and to evaluate the suitability of materials for release to unrestricted use.

Instrumentation and equipment will be capable of measuring the range of dose rates and radioactivity concentrations expected to be encountered during the decontamination and decommissioning activities associated with the BMRC, including implementation of a final status survey.

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BMRC Decommissioning Plan Revision 0 Decommissioning HP staff will select instrumentation that is sensitive to the minimum detection limits for the particular task being performed, but also with sufficient range to ensure that the full spectrum of anticipated conditions for a task or survey can be met by the instrumentation in use.

Consumable supplies will conform to manufacturer and/or regulatory recommendation to ensure that measurements meet desired sensitivity and are valid for the intended purpose.

3.1.2.4 Storage, Calibration, Testing, and Maintenance of Health Physics Equipment and Instrumentation Survey instruments will be stored in a common location under the control of BMRC Decommissioning HP Staff. A program to identify and remove from service inoperable or out-of-calibration instruments or equipment as described in HP procedures will be adhered to throughout the BMRC Decommissioning Project. Survey instruments, counting equipment, air samplers, air monitors and personnel contamination monitors will be calibrated at license-required intervals, manufacturer-prescribed intervals (if shorter frequency) or prior to use against standards that are traceable to the National Institute of Science and Technology (NIST) in accordance with approved calibration laboratory procedures, HP procedures, or vendor technical manuals. Survey instruments will be operationally checked daily when in use. Counting equipment operability will be verified daily when in use. The personnel contamination monitors are operationally tested on a daily basis when work is being performed.

3.1.2.5 Specific Health Physics Equipment and Instrumentation Use and Capabilities Table 3-2 provides details of typical HP equipment and instrumentation planned for use in the BMRC Decommissioning Project. This list is neither inclusive nor exclusive.

Table 3-2 Health Physics Instrumentation Radiation Calibration Instrument Detector Type Use Detected Source Ludlum Model 43-68 Gas Surface Static Measurements; Ludlum Model 2221 Beta Tc-99 Proportional (126 cm2 area) Beta scan measurements Ludlum Model 43-68 Gas Alpha/Beta static Ludlum Model 2360 Alpha/Beta Th-230/Tc-99 Proportional (126 cm2 area) measurements Ludlum Model 19 Internal NaI Gamma Cs-137 General area exposure rates Ludlum Model 43-10-1 ZnS Ludlum Model 3030E Alpha/Beta Th-230/Tc-99 Swipe/smear counting internal detector 50

BMRC Decommissioning Plan Revision 0 3.1.2.6 Potential Sources of Radiation or Contamination Exposure to Workers and Public as a Result of Decommissioning Activities Sources of radiation or contamination exposure were assessed by process knowledge, the HSA, radiological survey data, surveys performed during characterization, previous and current job coverage surveys, or daily, weekly and monthly routine surveys.

Worker exposure to significant external deep-dose radiation fields is possible during this project due to the level of activation of reactor components, the reactor bioshield, and the level of radioactive materials in the waste tank systems. Therefore, work precautions and ALARA techniques such as time, distance, shielding, and pre-job planning will be employed. Worker exposure to airborne radioactivity may occur during decontamination operations/work evolutions involving abrasives or methods that volatilize loose and/or fixed contamination.

Exposure of the public to external or internal radiation from this Decommissioning Project is possible due to the high dose rates of reactor components and the position of the Tank Farm, however, exposures to the public shall be kept below the limits specified in 10 CFR 20 and will be ALARA by using the Administrative and Engineering controls described in the following paragraphs.

The types of exposure controls used will consider the current state of technology and the economics of improvements in relation to the benefits. Control of potential sources of radiation exposure to workers and public as a result of decommissioning activities will be achieved through, but not limited to, the use of administrative and engineering controls.

Administrative controls consist of, but are not limited to:

x Administrative dose limits that are lower than regulatory limits x Training x Radiological surveys x Physical barriers such as radiological warning rope/ribbon, in combination with radiological warning tape, lockable doors/gates as well as information signs and flashing lights or other applicable barriers may also be used Engineering controls may consist of but are not limited to:

x HEPA ventilation/enclosures x Protective clothing/equipment x Containment(s) 51

BMRC Decommissioning Plan Revision 0 3.1.2.7 Health Physics Policies for Contractor Personnel Contractor personnel will be used during the BMRC Decommissioning Project and will be required, at the discretion of the RSO, to:

x Attend and complete appropriate radiation safety course x Provide required exposure history information x Read and sign an applicable RWP and comply with instructions x Follow special instructions given by the HP 3.1.3 Dose Estimates The total estimated occupational exposure to complete the BMRC Decommissioning Project is 6.7 person-rem. The dose estimate for decommissioning of the reactor was prepared using the individual work activity durations and work crew sizes, based upon the results of the characterization results to date and based upon recent experience in performing similar activities at the University of Washington and at the University of Arizona, combined with the DOCs experience at numerous other sites. Using these individual work activity durations, and work crew sizes and characterizations results, a dose estimate was generated for each activity. The doses from each activity were categorized and are provided by those categories in Table 3-3.

This estimate is provided for planning purposes only. Detailed exposure estimates and exposure controls will be developed in accordance with the requirements of the ALARA program during detailed planning of the decommissioning activities.

The primary doses expected to be received by D&D project workers will be from external exposure to activated metals and concrete, with little dose expected from internal exposure. External exposure will be monitored using whole-body and extremity TLDs, and electronic dosimeters. External exposures can be kept ALARA due to the availability of long reach tools, remote handling equipment, and the buildings overhead crane. Additionally, there is plenty of floor space in the BMRC to maintain safe distances and to use shielding as necessary to keep exposures ALARA. Air sampling will be performed to assess the potential for airborne contaminants and internal doses will be monitored if they are expected to exceed 10% of the annual dose limits specified in 10 CFR 20. However, the committed effective dose equivalent (CEDE), the sum of the external and internal doses, is expected to be equal to the Deep Dose Equivalent (DDE).

The dose estimate to members of the public as a result of decommissioning activities is estimated to be negligible. This is because the area immediately surrounding the facility is under the control of the UB and because the BMRC is not within a high foot-traffic area of the UB South Campus. UB students, 52

BMRC Decommissioning Plan Revision 0 faculty and staff will be kept at a safe distance from the BMRC during decommissioning activities using temporary construction fencing.

Table 3-3 Project Dose Estimate Time/Resources Est. Dose Task Task Name Days People (person -

rem) 1 Project Planning 40 7 0.04 2 Site Mobilization and Training 5 10 0.00 4 Removal of the Operating Platform 5 6 0.01 5 Removal of Reactor Components 15 8 2.88 6 Water Removal from the Reactor Tank 5 8 0.02 8 Removal of Reactor Bio-Shield, Dry Cell, and Hot Cell 80 8 2.56 9 Removal of HVAC and other impacted systems 50 8 0.16 10 Tank Farm Removal 30 8 0.96 11 D&D of Containment Building 20 10 0.02 12 D&D of Administration Building 20 10 0.02 13 Remediation of Soils and Bedrock 20 10 0.02 14 Perform Final Status Survey 20 9 0.01 15 Backfill and Site Restoration 20 8 0.00 Total Estimated Project Dose 6.70 3.2 Waste Management Decommissioning will require the handling of a relatively large volume of radioactive materials to reduce the residual levels of radioactivity to levels that allow for license termination and the release of the site for unrestricted use. Materials that are not decontaminated and/or released will be managed as radioactive waste. This section of the decommissioning plan presents the programs used to manage and control the management and disposition of solid, liquid and gaseous radioactive waste.

Radioactive wastes include neutron-activated materials, contaminated materials remaining in the containment building, tools, and disposable equipment and supplies that become contaminated during dismantling activities. Waste disposal costs are directly related to the activity, volume, and weight of the materials requiring disposal. Strategies for minimizing waste include: source reduction, reuse, decontamination, volume reduction, and waste stream segregation.

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BMRC Decommissioning Plan Revision 0 The DOC will be required to develop and implement a Waste Management Plan for the BMRC Decommissioning project. The Waste Management Plan will include detailed guidance for the characterization, sampling, classification, segregation, handling, packaging, manifesting, transporting and disposal of waste generated by the decommissioning. The plan will be submitted to the RSO and BMRC Director for review and approval prior to the start of any decommissioning work on site that has the potential to generate radioactive waste.

Waste volumes will be minimized through the application of industry-proven methodologies to ensure the segregations of contaminated and non-contaminated materials. These methodologies will include the establishment of radiological controls consistent with the health physics program and the implementation of good contamination control practices. Segregation categories may include: uncontaminated waste acceptable for land disposal or reuse, uncontaminated C&D wastes suitable for land disposal or recycle, Class A LLRW, and Class B activated components. Additionally, mixed wastes and non-radiological hazardous waste, if encountered, will be segregated from LLRW. Based on the site characterization and Activation Analysis of the reactor components, Class C radwaste is not present at the BMRC.

Except for Class B waste, disposal facility options for contaminated or activated materials currently available to BMRC include land disposal at the Energy Solutions facility in Clive, Utah; and mixed waste treatment by Energy Solutions or another qualified service provider, with subsequent disposal at Energy Solutions. BSFR using a bulk survey and disposal procedure that is specially-permitted in Tennessee is also an option for lightly contaminated materials, especially building materials.

Currently, there is no commercial Class B disposal option available for the Old Control Blades. The UB is working with the Nevada National Security Site (N2S2) to determine if a nexus exists between BMRC operations and the federal government in order to establish this as a disposal option. The UB shall properly manage the Class B waste until a disposal option is obtained.

Transportation of radioactive waste will be performed in accordance with applicable NRC and DOT regulations and the Waste Management Plan. Radioactive waste and material will be shipped either by truck, or by a combination of truck and rail. Shipments will be planned in a practical and efficient manner. Personnel with appropriate training and experience will be used to ensure the shipments comply with the BMRC license, applicable regulations, and the receiving sites license. Packages, packaging, and labeling for radioactive materials and waste shipment will meet applicable regulations and requirements.

Personnel involved in the packaging, preparation for shipment, and transportation of licensed material 54

BMRC Decommissioning Plan Revision 0 will be required to have training in accordance with 49 CFR 172 Subpart H. The RSO, or designee, shall certify all radioactive waste manifests.

3.2.1 Fuel Removal There is no longer any reactor fuel, used or unused, at the BMRC. As previously stated, the reactor was permanently shutdown in June of 1994. Unused fuel was transferred to North Carolina State University in 1998, and used fuel was returned to the DOE at INEEL in 2005.

3.2.2 Radioactive Waste Processing Decommissioning of the BMRC reactor will result in the generation of solid and liquid LLRW, mixed waste, and hazardous waste. Solid radioactive wastes include neutron-activated materials, materials with surface or volumetric residual radioactive materials, and soil. Liquid LLRW includes the water in the reactor pool, waste systems, and the associated piping as well as contaminated water generated during remediation activities. The DP does not contain provisions for gaseous radioactive waste because the reactor has been defueled and has been shut down for over 16 years; therefore, radioactive gases present during operations have long since decayed.

Handling, staging, and shipping of packaged radioactive waste will be performed in accordance with applicable regulations, 10 CFR 20.2006, Transfer for Disposal and Manifests; 49 CFR 100-177, Transportation of Hazardous Materials; disposal site waste acceptance criteria; and BMRC licenses and permits. Onsite radioactive waste processing will include waste minimization, volume reduction, segregation, characterization, neutralization, stabilization, solidification, and packaging. Wastes may be shipped to a licensed processing facility for survey and release or decontamination and release, or may be disposed of directly at a licensed facility. Each shipment of radioactive waste will be accompanied by a shipment manifest as specified in Section I of Appendix G to 10 CFR 20, Requirements for Transfers of Low-Level Waste Intended for Disposal at Licensed Land Facilities and Manifests. Radioactive waste generated from BMRC decommissioning activities will be manifest in a manner consistent with its waste classification.

3.2.3 Radioactive Waste Disposal 3.2.3.1 Solid Solid waste generation will primarily be the direct result of the decontamination and dismantlement of the BMRC reactor components, activated and contaminated systems, and structures. The bulk of the radwaste is Class A with approximately 2 cubic feet of Class B waste and no Class C wastes based on site 55

BMRC Decommissioning Plan Revision 0 characterization data and activation analysis data for the reactor. Disposition paths for solid radioactive wastes include decontamination and free release, BSFR, and direct landfill disposal. Wherever possible, volume reduction strategies will be explored to reduce waste disposal handling, exposure, and cost; including offsite processing for volume reduction if appropriate.

The Class B Mixed Waste consists of the Old Control Blades made of nickel plated permalloy which is an alloy consisting of silver, indium and cadmium. Disposal sites licensed to accepted Class B waste are not available to the BMRC at the time this DP was developed. The UB and DOC are initiating the process with the N2S2 to determine if the Class B waste can be accepted for disposal. After determination is made for disposal at N2S2, or another site, the Waste Management Plan is to be revised to incorporate the waste acceptance criteria for the disposal site.

Irradiated reactor hardware may require size reduction to facilitate loading. Depending on the dose rate exhibited by irradiated hardware, it will be shipped either in a shielded Type A container, or it will be loaded into a High Integrity Container (HIC) or liner then placed in an approved, shielded shipping container (Type A or B as dictated by the waste characterization) for transport and subsequent disposal.

Contaminated reactor system piping and hardware may be land disposed as LLRW. Pre-disposal volume reduction processing may be performed off site at a subcontracted processing facility if it is determined to be cost effective.

Activated or contaminated concrete removed in large sections may be packaged as Low Specific Activity (LSA) material in approved shipping containers for direct shipment to the licensed land disposal facility operated by Energy Solutions at Clive, Utah. Concrete may also be eligible for processing using an option such as the BSFR program offered in Tennessee, or potentially N2S2.

Mixed waste lead bricks will either be decontaminated and released, or processed to allow for direct land disposal.

3.2.3.2 Liquid Decommissioning the BMRC reactor involves radioactively contaminated water primarily in the form of low-level radioactively contaminated water in the reactor pool and associated piping. Additional contaminated water may be generated during decommissioning operations (e.g. concrete cutting). These waters will be disposed by discharge to the public sewer system operated by the City of Buffalo; assuming the discharged liquid can be shown to meet the requirements for sewage disposal established by the BMRC license.

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BMRC Decommissioning Plan Revision 0 Pre-discharge treatment may include the use of existing or temporary filtration units or demineralizers, coupled with tanks to store processed water prior to discharge. After it has been verified that the stored processed water meets the allowable discharge limits, the water may be released. In addition to testing the stored water after treatment, effluent monitoring instrumentation may be used to monitor discharges of liquid effluent as required, and to demonstrate compliance with applicable regulations. Contaminated filter media will be disposed as LLRW, after onsite treatment to meet LLRW disposal facility waste acceptance criteria.

In the event that discharge to the sanitary sewer or onsite treatment is not feasible, the waters may be treated offsite. There are several licensed radioactive waste processors provide specialized services for volume-reducing or treating radioactive liquid waste, including demineralization, direct incineration, ground application, evaporation, and survey and release. BMRC may elect, or find it necessary, to transfer all or some of the liquid radioactive waste from decommissioning to a licensed waste processor based upon onsite treatment effectiveness, discharge permit limits, stakeholder input, or cost.

3.2.4 General Industrial Safety Program Industrial safety and hygiene personnel, such as Certified Safety Professionals or Certified Industrial Hygienists, along with project management personnel, will be responsible for ensuring that the D&D project complies with applicable federal safety requirements and general safe work practices. These personnel will be provided by the DOC, but will report to the UB EH&S department. The DOC will prepare a site specific Health and Safety Plan (HASP) to document safety requirements and accident response procedures.

All personnel working on the D&D project will receive health and safety training in order to recognize and understand potential hazards and risks. Training requirements for subcontractors will be determined by the Decommissioning Safety Manager based on the specific task the subcontractor is performing.

The HASP will be reviewed and approved by the UB EH&S Department. The HASP will direct site activities necessary for ensuring that the reactor D&D project meets occupational safety and health requirements for protection of project personnel. The functional responsibility of the HASP will be to ensure compliance with the OSH Act of 1973. New York adopts federal OSHA standards by reference and enforces OSHA standards contained within 29 CFR 1910 and 29 CFR 1926 for General Industry and Construction Industry, respectively. The DOC assists the UB in oversight of the site HASP.

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BMRC Decommissioning Plan Revision 0 As a minimum, the HASP will include the following:

x Hazards assessment x General site safety procedures x A requirement for a daily site safety meeting x Site inspection procedures x Emergency response procedures x Emergency contact telephone numbers x Material Safety Data Sheets for hazardous materials present on-site x Training requirements for specific activities such as permit-required confined space entry or hot work x Local emergency medical information 3.3 Radiological Accident Analyses Potential radiological accidents during reactor decommissioning were evaluated by determining reactor components and areas that contain the highest radioactive material inventory. The proposed decommissioning activities and methods in which radioactive material could be released to the work area or environment were considered. Since all special nuclear material has been removed decommissioning, the majority of the accidents discussed in the current license are not applicable. The accident identification process was supplemented by reviewing experiences at other non-power reactor decommissioning projects. The following radiological accidents were considered to present the highest potential consequences:

x Fire x Spill liquid radwaste into the environment x Release airborne contamination to the environment x Transportation accident The accidental dropping of an activated reactor component was also considered as a potential accident.

However, because the more highly activated components are located under water, the surface contamination on these parts would not be sufficiently high to release significant quantities of radioactive materials during such an incident. Such an incident would most likely result in additional unplanned external exposures. There are no, and will not be, fissile materials located on site that could result in a criticality incident.

3.3.1 Fire The consequences of a fire during decommissioning of the reactor were considered and are not significantly different than the consequences of a fire during reactor operations. Most materials are metals, concrete, or similar non-combustible materials. Although some torch cutting operations may be 58

BMRC Decommissioning Plan Revision 0 performed during decommissioning, the likelihood is low that a fire would start or that a fire could become intense enough to release radioactive material.

Dry Active Waste (DAW) will be collected and packaged, to limit the volume of DAW available for consumption by fire and lower the potential for a fire to consume additional waste collections. The UB will implement a routine fire loading inspection program during decommissioning activities. DAW would have very low quantities of radioactive materials; therefore, the radioactivity is not high enough to result in a significant release in the case of a fire.

3.3.2 Spill Contaminated Water The consequences of a spill during decommissioning of the reactor were considered and are not significantly different than the consequences of a spill during reactor operations. The spilling of contaminated water could occur during pool water pumping or liquid removal operations from the waste systems. Hoses could leak or break, resulting in an uncontrolled release. To mitigate the extent of such releases, processes involving contaminated liquids will only be operated with personnel present.

Personnel will watch for leaks and spills and respond by shutting down the activity. This will not allow for additional water to leak from the system. In addition to the use of secondary containments or berms, a spill kit will be readily available to respond to any incidents.

As evidenced in prior reactor leaks, contaminated water does not infiltrate the subsurface soils or bedrock.

Should radioactive liquids be spilled, the soils and/or building materials can be remediated with little to no impact to the public or the environment.

3.3.3 Release of Airborne Contamination The consequences of an airborne contamination event during decommissioning of the reactor were considered and are not significantly different than the consequences of an airborne contamination event during reactor operations. An uncontrolled release of airborne radioactivity could occur during cutting and demolition activities involving contaminated or activated materials, such as removal and segmentation of reactor components, or removal of tank steel and concrete. The primary method of cutting the activated bioshield is a wet process; therefore, eliminating most if not all of the airborne hazard. Such activities may take place inside temporary containment structures equipped with local HEPA filter ventilation systems. Additionally, non abrasive cutting methods, i.e., hydraulic shears, will be used where possible to limit abrasive dusts and/or activated metal fragments where feasible.

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BMRC Decommissioning Plan Revision 0 Temporary containment systems with local HEPA filter systems will likely vent to the BMRC rooms or tie into existing building ventilation. A failure in the HEPA filter system could result in the uncontrolled release of airborne radioactive materials. A Continuous Air Monitor will be used to monitor effluent air and will be set to alarm at 10% of the allowable effluent criteria. Operations inside the containment structure will immediately stop and an evaluation conducted to determine the nature of the alarm.

While the actual concentrations of airborne radioactive materials are unknown at this time, the scenario is similar to accident analyses contained in the current BMRC Technical Specifications. Safety management operations (standard engineering and administrative controls) are sufficient for protecting against such accidents.

3.3.4 Transportation Accidents Various forms and quantities of radioactive waste will be shipped from the reactor during the D&D project. The dose consequence from transportation accidents could be higher than the contamination accident scenarios described above because high-activity reactor components could be involved. As such, there is a potential for a moderate dose consequence of between 1 and 25 mrem for the public following a transportation accident. However, adherence to NRC and DOT radioactive material packaging and transportation requirements is considered a sufficient control measure for mitigating transportation-related incidents. The transportation accident risk is further reduced by using local disposition methods for the concrete, i.e., a local C&D landfill or recycling.

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BMRC Decommissioning Plan Revision 0 4.0 PROPOSED FINAL STATUS SURVEY PLAN An FSS Plan will be submitted to the NRC for approval following submittal of this DP and prior to implementation of the FSS. The following sections describe the methodology and QA requirements to be implemented during the FSS. The DOC is responsible for the planning and implementation of the FSS.

4.1 Survey and Sampling Approach The reactor and support facilities will be removed prior to site release. Consequently, the Final Status Survey (FSS) will include only the exposed soils and bedrock surfaces in areas covered by license R-77.

The expected condition of the area post remediation will be exposed subsurface soils and bedrock. The non-porous bedrock, being similar to a concrete structure, will be surveyed using survey methodologies for surface release of concrete and the results compared to the NRC surface screening values. Prior to the FSS on bedrock material, the loose dirt and debris is to be removed to ensure there is no interference with the survey measurements. The BMRC Site Characterization report (ENERCON 2011) demonstrates that the subsurface soils are not volumetrically contaminated; therefore, the exposed subsurface soils will be surveyed to surface soil screening criteria to prove that the surface of these exposed soils were not contaminated during the building demolition process. There will be no BMRC buildings or structures that will require a FSS.

The FSS will be developed following the guidance provided in NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) (NRC 2000) to demonstrate compliance with the release criteria provided in Section 2.2.3. The MARSSIM process emphasizes the use of data quality objectives (DQO), proper classification of survey areas (survey units), a statistically-based survey and sampling plan, and an adequate quality assurance/quality control (QA/QC) program.

The FSS will be performed in accordance with an FSS Plan by trained DOC technicians experienced in performing a MARSSIM FSS. The technicians will follow written procedures regarding surveys and sampling, sample collection and handling, chain-of-custody, and recordkeeping. The FSS Plan will define sampling locations, required analysis, and survey types. Any additional release criteria set forth by the UB or NYS will be contained within the FSS Plan which will direct surveys or sampling efforts required to demonstrate compliance with such criteria.

The FSS may include surface gamma surveys using sodium-iodide (NaI) gamma scintillation detectors and gas flow proportional detectors. Surface soil samples will be collected using either a random-start grid pattern or randomly generated locations as appropriate commensurate to the classification of the survey area. Soil samples will be analyzed for contaminants of concern using standard analytical methods 61

BMRC Decommissioning Plan Revision 0 including liquid scintillation counting for hard-to-detect beta-emitting radionuclides (i.e., Carbon-14 and tritium) and gamma spectroscopy for gamma-emitting radionuclides.

4.2 Data Quality Objectives The object of the FSS is to demonstrate that the radiological conditions of the reactor site satisfy the decommissioning criteria provided in Section 2.2.3. The DQOs in the MARSSIM survey approach will provide a 95% confidence level for the false negative (Type I error) in demonstrating that the site meets the criteria. Typically, the false positive (Type II error) will also be defined as a 95% confidence level, but may be modified to apply to a specific situation. Therefore, the Type I decision error will be 5-percent. The decision error rates are used in determining the required number of samples necessary in each survey unit as well as the required minimum number of data points used for the final nonparametric statistical test performed to evaluate contaminant concentrations in the survey units against release criteria. DQOs, will be fully described in the FSS Plan and will include limits on the sensitivities of survey and analytical methods.

The QAPP will incorporate standard regulatory and industry measures applicable to the FSS. The QAPP will be reviewed and approved by the Reactor Committee.

4.3 Identification and Classification of Survey Units 4.3.1 Method for Classification Survey units are classified based on contamination potential according to the methods described in the MARSSIM. In general, there are two overall classifications, non-impacted and impacted. Non-Impacted areas have no reasonable potential for residual contamination because there was no known impact from facility operations.

Impacted areas may contain residual radioactivity from facility operations. Based on the levels of residual radioactivity present, impacted areas are further divided into Class 1, Class 2 or Class 3 designations.

Class 1 areas have the greatest potential for residual activity while Class 3 areas have the least potential for impacted areas. Each classification will typically be bounded by areas classified one step lower to provide a buffer zone around the higher class. Exceptions occur when an area is surrounded by a significant physical barrier, such as a wall, that would make transport of residual activity unlikely from one area to the adjacent area. In such cases, each area will be classified solely on its own merit using the most reliable information available. The class definitions provided below are from Section 4.4 of the MARSSIM.

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BMRC Decommissioning Plan Revision 0 Class 1 Areas that have, or had prior to remediation, a potential for radioactive contamination (based on site operating history) or known contamination (based on previous radiological surveys). Examples of Class 1 areas include: 1) site areas previously subjected to remedial actions, 2) locations where leaks or spills are known to have occurred, 3) former burial or disposal sites, 4) waste storage sites, and 5) areas with contaminants in discrete solid pieces of material high specific activity. Note that areas containing contamination in excess of the DCGL w prior to remediation should be classified as Class 1 areas.

Class 2 These areas have, or had prior to remediation, a potential for radioactive contamination or known contamination, but are not expected to exceed the DCGL w .

To justify changing an area's classification from Class 1 to Class 2, the existing data (from the HSA, scoping surveys, or characterization surveys) should provide a high degree of confidence that no individual measurement would exceed the DCGL w .

Other justifications for this change in an area's classification may be appropriate based on the outcome of the DQO process. Examples of areas that might be classified as Class 2 for the final status survey include: 1) locations where radioactive materials were present in an unsealed form (e.g., process facilities), 2) potentially contaminated transport routes, 3) areas downwind from stack release points, 4) upper walls and ceilings of some buildings or rooms subjected to airborne radioactivity, 5) areas where low concentrations of radioactive materials were handled, and 6) areas on the perimeter of former contamination control areas.

Class 3 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 DCGL w ,

based on site operating history and previous radiological surveys. Examples of areas that might be classified as Class 3 include buffer zones around Class 1 or Class 2 areas, and areas with very low potential for residual contamination but insufficient information to justify a non-impacted classification.

The size of a survey unit is directly affected by its classification. Section 4.6 of MARSSIM provides suggested sizes for survey units. However, as stated in MARSSIM, the suggested survey unit sizes were based on a finding of reasonable sample density and consistency with commonly used dose modeling codes. MARSSIM limits the size of a survey unit. Table 4-1 summarizes MARSSIM recommendations for survey unit sizes based on their type and classification.

Table 4-1 Recommended Survey Unit Sizes Minimum/Maximum Classification Buildings Open Land 2 2 Class 1 10-m / 100-m 100-m2 / 2,000-m2 Class 2 100-m2 / 1,000-m2 2,000-m2 / 10,000-m2 Class 3 1,000-m2 / No limit 10,000-m2 / No limit 63

BMRC Decommissioning Plan Revision 0 4.3.2 BMRC MARSSIM Classifications Following remediation and demolition of the buildings, there will be no structures remaining that will require FSS. Using the size restrictions in Table 4-1, the exposed bedrock may include only one Class 1 survey unit. The exposed subsurface soils surrounding the Class 1 survey unit will likely be designated as a single Class 2 survey unit to a maximum survey unit size of 10,000 m2. The remaining surface soils will be designated as a Class 3 survey unit.

4.4 Data Collection Survey methods are applied differently depending on the data requirements of a survey area. For example, removable activity measurements provide little, if any, benefit when attempting to assess the radiological conditions in an excavation. Conversely, assessing a building surface via volumetric sampling would provide the necessary data, but at great costs of time and money. This section will discuss the steps necessary to strike a reasonable balance between data needs and ease of survey performance based on the data needs of the survey area.

4.4.1 Buildings, Equipment, and Components Buildings, equipment, and components are not expected to remain after license termination.

4.4.2 Soils To demonstrate soil areas meet the appropriate release criteria, gamma scans and volumetric samples are required as demonstrated in the following two sections.

4.4.2.1 Surface Soil Scans Soil areas require gamma scan measurements as part of the FSS process at appropriate coverage rates and speeds that ensure hot spots and/or hot particles are detected. The coverage rates and speeds will be described in the FSS plan and subsequent report to ensure adequate MDCs for hot spots and/or particles.

4.4.2.2 Volumetric Samples Volumetric samples are required to demonstrate a soil area meets the appropriate release criteria. In lieu of volumetric samples, soil areas may receive direct measurements using in-situ gamma spectroscopy, as equipment and trained personnel are available. Volumetric sampling differs slightly depending on the situation for which the sample is desired. The required quantity of volumetric samples for an open land survey unit is a calculated value that is discussed in MARSSIM.

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BMRC Decommissioning Plan Revision 0 4.4.3 Bedrock The Neutron Deck of the BMRC was poured directly on the bedrock with no leveling material (e.g.,

gravel or sand) utilized between the bedrock and concrete. The bedrock, described in Section 8.0, is non-porous and similar to a concrete slab. Radiological impacts to the bedrock are expected to be limited to the surface and after the concrete is removed from the bedrock, loose material such as soil and concrete dust will be removed prior to the FSS.

4.4.4 Bedrock Surface Scans The bedrock surfaces, after cleaning, will be scanned using beta instrumentation such as a gas flow proportional detector. The coverage rates and speeds will be described in the FSS plan and subsequent report to ensure adequate MDCs for hot spots and/or particles. Volumetric samples may be collected to verify residual radioactive materials are only on the surface of the bedrock.

4.5 Data Evaluation Data evaluation is performed on FSS results for individual survey units to determine whether the survey unit meets the release criteria. Appropriate tests will be used for the statistical evaluation of survey data.

Tests such as the Sign test and Wilcoxon Rank-Sum (WRS) test will be implemented using unity rules, surrogate methodologies, or combinations of unity rules and surrogate methodologies, as described in the MARSSIM and NUREG-1505 Chapters 11 and 12.

If the contaminant is not in the background or constitutes a small fraction of the DCGL, the Sign test will be used. If background is a significant fraction of the DCGL the WRS test will be used. It is anticipated that the sign test will be the only statistical test applied to the collected data because of the small fraction of the DCGL that background radionuclides will contribute.

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BMRC Decommissioning Plan Revision 0 4.6 Final Status Survey Report The FSS report will provide a summary of the survey results and the overall conclusions to demonstrate that the BMRC site meets the radiological criteria for release. Information such as the number and type of measurements, basic statistical quantities, and statistical analysis results will be included in the report.

The level of detail is to be sufficient enough to clearly describe the FSS program and to certify the results.

The basic outline of the final report will be similar to the following:

x Overview of the Results x Discussion of Changes to FSS x FSS Methodology o Survey unit sample size o Justification for sample size o Survey Instrument MDCs x FSS Results o Number of measurements taken o Survey maps o Sample concentrations o Statistical evaluations, including power curves o Judgmental and miscellaneous data sets o Elevated Measurement Comparisons (if used) x Conclusion for each survey unit 66

BMRC Decommissioning Plan Revision 0 5.0 TECHNICAL SPECIFICATIONS This DP is an addendum to Revision 27 of the BMRC Technical Specifications. Revision 27 has been submitted to the NRC via a license amendment request to perform decommissioning activities in accordance with this plan. The Technical Specifications include a license condition requiring an FSS Plan to be submitted and approved by the NRC prior to conduct of license termination surveys.

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BMRC Decommissioning Plan Revision 0 6.0 PHYSICAL SECURITY PLAN The regulations in Section 73.67(c)(1) of Part 73 require facilities to maintain a physical security plan when they possess special nuclear materials of moderate strategic significance or 10 kg or more of special nuclear material of low strategic significance. There is no nuclear fuel remaining at the BMRC; therefore, a part 73 physical security plan is not required.

A site security plan is maintained to comply with the regulations in Subpart I, Storage and Control of Licensed Material, of 10 CFR 20 are applicable to the remaining radioactive materials possessed by the UB. Licensed materials in storage are secured from unauthorized access or removal. Additionally, licensed materials not in storage will be under the control and constant surveillance of authorized personnel.

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BMRC Decommissioning Plan Revision 0 7.0 EMERGENCY PLAN Overview BMRC ceased operation in 1994, with the used reactor fuel returned to the DOE in 2005 and unused reactor fuel shipped in 1998. Current activities in this building are minimal, with concentration on maintaining surveillance, monitoring of environmental conditions and security of the facility. During decommissioning of the site, management of significant emergency incidents by the UB will be accomplished through implementation of a response framework as specified by Federal Emergency Management Agency (FEMA) using the Incident Command System (ICS). The UB is committed to an All Hazard Emergency Management approach, which designates areas of responsibility and defines the administrative framework to respond to all emergency incidents.

Actions The start of the decommissioning work raises the level of concern regarding response to an incident at the BMRC. Even though research has ended, the fuel returned to the DOE, and the level of risk reduced, the importance of protecting the surrounding community remains a priority to the UB.

Without research activity or a source of fuel, the existing equipment is the highest risk at the BMRC.

Access to the facility is limited and daily monitoring of activities at the BMRC is conducted under the supervision of the UB and DOC project management team, minimizing the risk potential. This team carefully monitors the status of activities and implements controls to minimize exposure. The Site Security Plan, along with the UB police monitoring capabilities, maintains a safe environment and protects the surrounding community.

The UB Communications Department plays an important role in response to any emergency at BMRC, providing accurate information to all media resources. Utilizing the Crisis Communication Plan from the Division of External Affairs, this information flow will maintain community confidence and supporting the UBs response to all incidences at the BMRC.

Utilization of the ICS, under FEMA guidelines, will provide a quick response mitigating the incidents of concern at the BMRC. Unified command under ICS format will afford the UB the opportunity to bring experts from multiple agencies together to mitigate impacts and allow for a smooth transition back to normal activities at the BMRC until decommissioning.

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BMRC Decommissioning Plan Revision 0 8.0 ENVIRONMENTAL REPORT The Decommissioning ER was prepared in accordance with the guidance provided in Chapter 6.0 of the NRC Office of Nuclear Material and Safety and Safeguards' (NMSS) NUREG-1748, Environmental Review Guidance for Licensing Actions Associated with NMSS Programs (NRC 2003b). The ER is provided in Appendix C and is provided for use by the NRC to conduct its environmental assessment of the potential impacts of this proposed DP in accordance with the National Environmental Policy Act (NEPA) of 1969. NEPA requires Federal agencies, as part of their decision-making process, to consider the environmental impacts of actions under their jurisdiction. The NRC's NEPA requirements are provided in 10 CFR 51.

70

BMRC Decommissioning Plan Revision 0 9.0 CHANGES TO THE DECOMMISSIONING PLAN Following NRC review and approval of the DP, the DP will be incorporated as an amendment to license R-77. Minor changes to the DP that do not change the original intent of the DP and which do not involve an unreviewed safety question may be approved by the Operating Committee, a sub-group of the RDSC, as defined in the BMRC Technical Specifications and this section. The licensee may make changes to the DP without prior NRC approval provided the proposed changes do not:

i. Require Commission approval pursuant to 10 CFR 50.59; ii. Use a statistical test other than the Sign test or WRS test for evaluation of the FSS; iii. Increase the radioactivity level, relative to the applicable derived concentration guideline level, at which an investigation occurs; iv. Reduce the coverage requirements for scan measurements;

v. Decrease an area classification (i.e., impacted to unimpacted; Class 1 to Class 2; Class 2 to Class 3; or Class 1 to Class 3);

vi. Increase the Type I decision error; vii. Increase the DCGLs and related MDCs (for both scan and fixed measurement methods); or viii. Result in significant environmental impacts not previously reviewed.

If a significant change to the DP is required, the BMRC RDSC will apply the criteria identified in 10 CFR 50.59 (March 2001) as it applies to non-power reactors in decommissioning. Guidance on implementing the requirements 10 CFR 50.59 is provided in the following documents:

x NRC Regulatory Guide 1.187 Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments x Nuclear Energy Institute (NEI) Guidance NEI 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1, September 2000 x NRC Inspection Guidance (Part 9900)

If the RDSC determines that the change is significant and could pose a significant increase in potential worker, public, or environmental impacts, NRC approval will be obtained prior to implementing the change.

Changes to the DP are to be listed in the Summary of Changes table at the beginning of this DP with special demarcations in the margins next to the revised text. A report of changes made to the DP without NRC approval is to be maintained for review by the NRC during routine decommissioning inspections.

Records of all changes to the DP are maintained until license termination.

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BMRC Decommissioning Plan Revision 0

10.0 REFERENCES

1. Enercon Services, Inc. July 30, 2011. BMRC Radiological Characterization Report.

2. WMG, Inc. September 2011. University at Buffalo Activation Analysis and Component Characterization, Revision 2.

3. International Organization for Standardization (ISO). 1988. Evaluation of Surface Contamination-Part 2: Tritium Surface Contamination. ISO 7503-2. August 1.

4. U.S. Nuclear Regulatory Commission (NRC) (formerly the U.S. Atomic Energy Commission).

1974. Termination of Operating Licenses for Nuclear Reactors. Regulatory Guide 1.86. June.

5. NRC. 1981. Control of Radioactively Contaminated Materials. NRC Office of Inspection and Enforcement (IE). IE Circular 81-07. May 14.

6. NRC. 1996. Guidance for Preparing and Reviewing Applications for Licensing of Non-Power Reactors.NUREG-1537, Part 1. February.

7. NRC. 2006. Consolidated Decommissioning Guidance. NUREG-1757

8. NRC. 2000. Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM). NUREG-1575, Revision 1. August.

9. NRC. 2003. Radiological Assessments for Clearance of Materials From Nuclear Facilities.

NUREG-1640, Volume 1. June.

10. NRC. 2003. Environmental Review Guidance for Licensing Actions Associated with NMSS Programs. Office of Nuclear Material and Safety and Safeguards. NUREG-1748

11. NRC. 2011. Standards for the protection against radiation Title 10, Part 20, Code of Federal Regulations.

72

BMRC Decommissioning Plan Revision 0 APPENDIX A - ACTIVATION ANALYSIS AND COMPONENT CHARACTERIZATION

The T STAATE UNIV VERSITY Y of NEW W YORK K at BUF FFALO BU UFFALO MATER RIALS RESEARC CH CENNTER REEACTORR ACCTIVATIION ANA ALYSIS AND COMPPONENT T CHAR RACTERIIZATION N Report 10-102D-RE 1 E-138 Reevision A May M 2011 WM MG Projectt 10-102D 1

Preepared by:

16 Bank B Stree et Peekskkill, NY 10566

FO OREWORD This rep port summaarizes the acctivation an nalysis workk performedd by WMG, Inc. to sup pport The Staate University of New York (SUN NY) in the preparationn of the deecommissiooning plan for the University at Bufffalos Reactor. This work was performed d by WMG,, Inc.

under SUNY Contrract T-000401.

BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Characcterization 05/2 2011 i

TABLE OF O CONTE ENTS 1.0 NTRODUCTION .................................................................................................. 1 IN 2.0 NEUTRON N TRANSPO T RT AND AC CTIVATION N METHOD DOLOGY ........................ 3 2.1 INPUTT PARAMETE ERS ............................. ........................................................ 4 2.2 NEUTRON FLUX DATA ................................................................................... 5 2.3 ORIGGEN 2 NEUTTRON ACTIVA ATION CALC ULATIONS ......................................... 7 3.0 ESTIMATED E D COMPON NENT RADIOACTIVIT TY .................................................... 8 3.1 OVERRVIEW ........................................... ........................................................ 8 3.2 COMP PONENT RAD DIOACTIVITY ................. ...................................................... 10 3.2.1 Reactor Structural S / Peripheral Componen nts ................................. 10 3.2.2 Reactor CoreC Comp ponents ..... ...................................................... 15 3.2.3 Experime ental Facilitiies .................................................................. 22 4.0 COMPONEN C NT PACKA AGING REQ QUIREMEN NTS ................................................ 27 5.0 REFERENC R CES ................................................................................................... 28 APPENDDICES Appendix A A - Origen Material M Com mpositions BMRC Activation A Analysis A and d Reporrt 10-102D--RE-138 Re ev. A Componnent Characcterization 05/2 2011 ii

LIST OF TABLE ES Table Title P Page 2-1 BMRC C Operating History ........................................................................... 5 3-1 Comp ponent Cha aracterizatioon Summarry ..................................................... 9 3-2 Plenuum Chambe er Characte erization Re esults ............................................. 11 3-3 Plenuum Chambe er Stand Ch haracteriza tion Resultts ................................... 12 3-4 Flappper Charactterization Results R ....... ...................................................... 13 3-5 Ion Chamber C Guuide Charac cterization Results.......................................... 14 3-6 Fissio on Chambe er Characte erization Re esults .............................................. 15 3-7 Grid Plate P Chara acterizationn Results ... ...................................................... 16 3-8 Grid Plate P Plug Characteriz C zation Resu ults ................................................. 17 3-9 Fissio on Chambe er Guide Ch haracterizattion Resultss ................................... 18 3-10 Contrrol Elementt Characterrization Ressults ............................................... 19 3-11 Contrrol Elementt Shroud Ch haracteriza tion Resultts ................................... 20 3-12 Contrrol Elementt Drive Shafft Characte erization Re esults............................. 21 3-13 Pneumatic Conv veyor Chara acterization n Results........................................ 22 3-14 Thermmal Column n Characterrization Ressults .............................................. 23 3-15 Isotoppe Tube Ch haracterizattion Resultss .................................................... 24 3-16 Vertic cal Standpip pe Charactterization R Results ............................................ 25 3-17 Isotoppe Tube Ch haracterizattion Resultss .................................................... 26 LIST OF O FIGURE ES Figure Title P Page 2-1 Univeersity of Bufffalo Irradia ation Facilityy Flux Data a ...................................... 6 BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/2 2011 iii

1.0 IN NTRODUCTION In Octob ber 2010, The T State University U of o New Yorkk at Buffalo o (SUNY Bu uffalo) enga aged WMG to o perform ana activation analysis and segme entation pla anning for the Universsitys Buffalo Materials Research Center (B BMRC) rea actor. Thiss report summarizess the preliminary results s of WMG's demonsttrated and approved analytical methodolo ogies used to complete this phase e of work. The prelim minary resu ults presennt the activation product content witthin the activated reac ctor compon nents.

The results presen nted in this report will require furrther norma alization aftter final surrveys have be een obtaine ed with ade equately sc caled meassurement e equipment. The repo orted results are a conside ered preliminary until all a (includin ng the conttrol rod shrouds and lower core plaate) the calculated acttivation activities are normalized d to measured dose rates, and surrface contamination, whichw cove ers the surffaces of th he reactor componentts, is evaluate ed.

Based on WMGs extensiv ve experie ence, the preliminarry activatio on results are conservative and provide a reasonab ble basis ffor decommissioning planning. In addition, surface contaminatio c y does not dictate the on typically e NRC waste classification of highly y activatedd reactor components c s. Howeve er, the surrface conta aminant acctivity (which can contaiin activatio on products s, fission products a and transuranics) can n be significa ant for reacctor compon nents with relatively low activatiion activitie es and musst be considerred prior too classificattion as Low w Level Rad dioactive W Waste (LLR RW) for disp posal in accordance with 10 CFR Pa art 61.

The rele evant activvation prod duct radion nuclide con ncentrationss determin ned from these results are a used to o classify the t activate ed reactor componen nts in accordance witth 10 CFR Pa art 61. The e activationn product scaling s facto ors are alsso presente ed in this re eport and are e used to quantify q thee hard-to-ddetect radio onuclide co oncentration ns important to final clas ssification under u 10 CFR Part 61. The repo orted scaling factors are Co-60 based and considered fina al since the e ratios of impurities inn the initial material coompositionss are assumed to be fixe ed and cons sistent thro oughout the e componen nts, as shown in Appe endix A. The erefore, thee individual radionuclide concen ntrations froom neutron n activation n are directly proportiona p al to the Co-60 concen ntration.

The ana alysis methoods used fo or this projeect are disccussed in S Section 2. T The compo onent specific results in n terms of o estimate ed activatio on activityy and 10 CFR Parrt 61 classification as of o June 1, 2011 arre presente ed in Secction 3. T The packa aging requiremments for thhe activated d componen nts are disccussed in S Section 4. Referencess are presente ed in Sectioon 5.

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Historiccal summary The BM MRC is a lig ght water pool p type reactor r con nstructed b by the Univversity that was originally y fueled witth Materials s Test Reactor (MTR)) fuel from 1 1961 to 196 63. The BM MRC was reco onfigured and a used AMF/Hittman n/IRM Pulsttar fuel fromm 1964 thro ough shutdown.

Active core height is app proximately 24-inchess and corre mid-pla ane is loccated approximmately 16-in nches abov ve the top face f of the Grid Plate.. Many rea actor core items were mo odified and re-used du uring the re econfiguratio on. The co ore is configgured to acccept a 6 x 6 array a of fue el assemblies but did not operate e with each h grid locatiion fueled. The BMRC operated o wiith relatively y high research reactoor flux levells approachhing 1013 n//cm2-

s. Watter cooling, experime ental facility y, and reaactor tank configuratiions were also changed d during the e operation o the facilitty without ssubstantial changes to nal history of o the reactor core c compo onents.

The facility operatinng history forms f the basis b e activation analysis p for the presented in n this document. Facility y operating records thrrough June 1994 are in ncluded in this report.

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2.0 NEUTRON N TRANSPO T RT AND ACTIVATIO A N METHOD DOLOGY WMG us sed Univerrsity supplie ed neutron flux levels at location ns of interest. A summary of the provided p eutron flux data is in ne ncluded in Section 2.2. The flu uxes and iinitial material compositioons shown in Appendiix A are use ed as inputts to the ORRIGEN2.2 (Ref.

5) compputer prograam to perforrm activatio on analysis on individu ual compon nents.

A similaar activatio on analysis methodo ology has been use ed by WM MG to sup pport decomm missioning activities at 18 do omestic po ower and research reactors. The methodo ology has been b refinedd and benc chmarked o over the yeaars and hass been foun nd to provide reasonable e characteriization resu ults for the ccomponentts of interesst.

Several assumptio ons were madem for th he activatioon analysiss, based up pon information providedd by the Unniversity or available from f simila performed by WMG in ar projects p n the past. Thhe activatio on analysis and dose profiling at the Brookh haven Natio onal Labora atory High Fluux Beam Reactor R indicated Co-60 producttion in 606 61 Aluminum m components.

Similar effects e to a lesser de egree are present p at tthe BMRC.. These assumptionss are stated in n the subs sequent sections describing eacch portion of the actiivation ana alysis methodo ology.

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2.1 In nput Param meters Use U of the ORIGEN2 2 computerr program requires a accurate m modeling off the re eactor com mponents in i terms of o their ph hysical cha aracteristicss and neu utron exposure his stories.

2.1.1 Material Compo ositions Elemental mate erial compo ositions are e required inputs for the ORIG GEN2 comp puter program. A su ummary of the initial material ccomposition ns is includded in the appendice es. Detaile ed drawings with matterial heat data were lacking, and a WMG used draw wings and technical description ns to discern materia als of cons struction. The Grid Plate and Control Rod Shrou uds are fabricated prrimarily of 1100 grad de Aluminum with inte egral stainless steel fasteners.

f Aluminum componen nts outsidee the core w were confirrmed or as ssumed to be fabrica ated from lless expen nsive and m more comm mon 6061 gradeg Alum minum. All stainless steel was assumed tto be type 304. Nick kel clad, silv ver-indium--cadmium C Control Ro od material was well documente ed. Reacttor grade graphite w was used in the the ermal colummn. The re eactor tankk is constru ucted of a 6061 Alum minum tankk and liner interior i to a poured 22 25 pcf concrrete pit.

The initial i mate erial compo osition dataa for SS304 4 and conccrete are taken from NUREG CR C 3474 (R Ref. 6). Th he graphite e composition is compiled from work perfo ormed for th he Brookha aven Nation nal Laborattories reseearch reactoors. The Aluminum A Alloys A 60611 and 1100 0 are comp piled from work performed at the e Universityy of Michig an and the e NASA Plu umbrook faccility.

Materrial compos sition data for f the Con ntrol Rod Blades was provided byy the Univeersity, 2.1.2 Operrating Histo ory A dettailed opera ating and power p historry was provvided by thhe Universitty for the BMRC.

B Thee informatioon was pro ovided in teerms of MW W-hrs produced on a monthly ba asis from sttart-up (Jully 1961) un ntil shutdow wn (June 19 994).

The cumulative c power prooduced ove er the lifetim me of the reactor thro ough June 1994 was 68,543.4 6 MW-hrs.

M An irrradiation po ower historry was com mpiled from m this informmation for each calenndar year of o operation n, which is considered d a cycle of operatio on in this analysis.

a The capacity y factor for each cycle e is determ mined using g the maxim mum MWs s/day (cycle 17, 19 987). Significant do own times are accou unted for in the op perating h history. TThe operating historry is summ marized in Table T 2-1 below.

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Table T 2-1 BMRC Op perating History Total OrigenPowwerHistory: "C Cycle" Deccay Year MWsper Capacity C

YearCycle Start Stop Leength Dayys MWs day** Factor 1961(712) 07/01/61 0 01/01/62 184 413.3 2.2 0.074 1962 01/01/62 0 01/01/63 365 1,295.6 3.55 0.118 1963 01/01/63 0 06/30/63 180 641 753.0 4.18 0.139 1965 04/01/65 0 01/01/66 275 3,255.8 11.84 0.392 1966 01/01/66 0 03/31/66 89 6 6485 1,085.3 12.19 0.404 1984 01/01/84 0 01/01/85 366 10,598.0 28.96 0.960 1985 01/01/85 0 01/01/86 365 10,922.7 29.93 0.992 1986 01/01/86 0 01/01/87 365 10,117.0 27.72 0.919 1987 01/01/87 0 01/01/88 365 11,010.4 30.17 1.000 1988 01/01/88 0 01/01/89 366 9,894.4 27.03 0.896 1989 01/01/89 11/30/89 333 914 9,025.8 27.10 0.899 1992 06/01/92 12/31/92 213 366 3,684.6 17.30 0.573 1994(13) 01/01/94 0 03/31/94 89 62 1,580.8 17.76 0.589 1994(6) 06/01/94 0 06/30/94 29 6 6083 0.8 0.03 0.001 Survey: 2/24/2011 Notes:

N x Time Steps (i.e., "Cycles") ") are estab blished bassed on the applicabiliity of actuaal operating g history at BMRC B vs O Origen com mputer code e x ** MW Ws/day is necessary due d to the in nconsistentt Cycle leng gths x Capa acity Factorr = [MWs/da ay ---- Cyclee(i)] / (Max.. MWs/day))

2.2 Neutron N Flu ux Data A summary of the neu utron flux data used in n the activa ation analyysis is showwn in Figure F 2-1 below.

b This s data represents typiical therma al and fast fflux data foor the irrradiation fa acilities.

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Figure F 2-1 Univers sity of Buffalo o Irradiation Facility Flux D Data B

BMRC Activatiion Analysis an nd Report 1 10-102D-RE-138 Rev. A C

Component Chharacterizationn 05/2011 Page 6 of 28

2.3 ORIGEN O 2 Neutron N Acctivation Calculation C ns The T ORIGEN2 computter code, ve ersion 2.2, was used to calculatte the activation and depletio on of radio onuclides inn compone ents expose ed to neutrron flux. E Each component was irradia ated based on the rea actor operatting history summarize ed in Table T 2-1 using u the appropriate e Universitty-supplied d flux for tthe compo onent lo ocation and d the initial material m compositions .

Some S reactor compon nent assem mblies, nota ably the Grrid Plate an nd Control Rod Shrouds, S arre construc cted from multiple mmaterials. Neutron acctivation differs significantly for each material and due to the differrent materia al compossition, reesulting in different dose rates and differe ent waste cclasses forr the respe ective component materials. However, the compo onents are treated as assemblie es for classificationn and deco ommissionin ng purposees in many cases due to the pracctical trradeoffs annd difficultiees in mate ation. For example, the Grid P erial separa Plate, primarily constructed of o 1100 grrade alumin num, includdes stainleess steel ddowel pins that would w be difficult d andd dose inteensive to remove. Therefore,, the classificationn results in n Section 3 for such h compone ent assemb blies reflectt the different material consttituents of thhe assembly.

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3.0 ESTIMATED E D COMPON NENT RAD DIOACTIVIT TY This secction presennts the estimated activ vity for eac h reactor ccomponent of interest a as of February y 2011. These T prelim minary estim mates are based on activation a analysis reesults and, as stated prev viously in th his report, will w be norm malized to mmeasured d dose rates w when such daata is availa able. The activities a prresented in this sectio on represen nt the activation productss only.

3.1 Overview O

The T activation results of all reac ctor compo onents pressented in tthis section n are based on ORIGEN O output o usingg Universitty provided d flux leve els and en nergy spectra. The ORIGE EN output is in termss of speciffic activity (Ci/g) and d the in ndividual nuuclide activ vities are caalculated ussing the acctivated com mponent weight and nuclide specific ac ctivity. Normalization n to more a accurately m measured dose ra ates will ch hange the activation a activity a for e each comp ponent but will not cha ange th he compon nent-specific scaling factors. B Based on WMGs pa ast experieence, normalizatioon to measured dose rates at th e time of a actual decoommissionin ng is expected to reduce com mponent total activity.

Table T 3-1 presents p a summary of the pre eliminary rresults in tterms of w waste weight, w wastte volume, total activa ation activityy and mateerials of connstruction. The estimated ac ctivity for all reactor coomponentss should be e less than 1,200 curie es as of June 1, 2011.

2 Com mponent waste w weigh hts, waste volumes and materia als of constructionn were dete ermined from the draw wings, docu uments and d data provvided by the Un niversity, supplement s ted as n necessary by WMGs enginee ering experience. Please no ote that thee waste vo lumes inclu uded in thiss section oof the analysis are e displacem ment volum mes and d o not refle ect the ass-packaged d for disposal, or envelope volumes. This will be e addressed d further in Section 4.00.

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TABLE T 3-1 University y at Buffalo BMRC B

Co omponent Ch haracterizationn Summary Waste Waste Total Component C Weight Volume Activity Materrials of Consttruction 3

(lbs) (ft ) (Ci)

Reactor Struc ctural/Periphera al Components Plenum 54.8 0.324 7.29E-01 6061 Aluminum m with 304 SS BBolting Flapper 19.0 0.113 2.27E-01 1100 Aluminum m Plenum Chamb ber Stand 31.9 0.190 4.41E+00 6061 Aluminum m Ion Chamber Guide G 7.2 0.043 1.00E+00 6061 Aluminum m Fission Chamb ber (worst case)) 1.5 0.009 1.18E+00 6061 Aluminum m with Uranium Loading Reactor Tank 6061 Aluminum m Subtotals <10 Reactor Core Components Grid Plate 88.4 0.522 7.00E+01 1100 Aluminum m with 36 SS Doowell Pins Grid Plug (ea) 2.3 0.014 1.67E+00 6061 Aluminum m Fission Chamb ber Guide 10.8 0.063 8.30E+00 6061 Aluminum m Control Elemen nt (ea, typical) 11.2 0.025 5.53E+01 Ni clad Ag-In-C Cd with Al & 304 4 SS fixtures Control Elemen nt Shroud (ea, typical) t 87.1 0.503 8.85E+01 6061 Aluminum m with SS Fasteeners Control Elemen nt Drive Shaft (eea, typical) 87.1 0.503 6.29E+01 6061 Aluminum m Subtotals <1,000 Experimental Facilities Pneumatic Con nveyor 10.3 0.061 1.43E+00 6061 Aluminum m Thermal Colum mn 3,162.0 22.0 1.16E+00 Rx Grade Grap phite & 6061 Aluminum Dry Chamber 6061 Aluminum m & Lead Shutteer Isotope Tubes (ea, typical) 1.8 0.011 1.33E+00 6061 Aluminum m Vertical Standp pipes (ea, typica al) 23.5 0.140 3.27E+00 6061 Aluminum m Becquerel Stan ndpipe 33.1 0.196 4.59E+00 6061 Aluminum m Subtotals <50 Totals <1100 B

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Component Ch haracterization n 05/2011 Page 9 of 28

3.2 Component C t Radioactivity In ndividual co omponent estimated e activation a a activities aree decay co orrected to June 1, 2011. The reactor componen nts are divided into th hree groupss: the Rea actor Structural/Pe S eripheral Components C s, the Expe erimental F Facilities and the Rea actor Core C Components. The T Reactor Core Components hold the fu uel elemen nts in place (Grid Plate) or re eside within n the reacttor core (Control Rod Shrouds, Core Plugs, P etc.)). Based on the em mpirical da ata from p previous de ecommissio oning activities, the e additionaal effect of surface s con ntaminants could incre ease compo onent dose rates marginally m but b is unlike ely to chang ge waste classification n.

The T radionuclide actiivities pres sented in this report are tho ose consid dered s

significant per the guidance prresented in n NUREG//BR-0204 Instructionss for Completing C NRCs Uniiform Low-L Level Radio oactive Wa aste Manife est (Ref. 7). In summary, a radionucliide is cons sidered sig gnificant wh hen it is coontained in n the waste w in conncentrations s:

x Greatter than 1%  % of the tota al activity off the compo onent x Greatter than 1%  % of the Cla ass A limit fo or nuclides listed in 100 CFCR 61.55 x Greatter than 7.0 0 PCi/cc for nuclides not listed in 10 CFR 61.55 As A discusse ed previously, final surrveys of the e reactor ccomponentss are necesssary o update the prelliminary results pre to esented h herein to provide final characteriza ation results s.

3.2.1 Reac ctor Structu ural/Periph heral Comp ponents The reactor r stru uctural and d periphera al compone ents include e the Conccrete Reac ctor Tank (Bioshield Concrete),

C Aluminum Reactor T Tank and L Liner, Plenu um and Flap pper, and Core C Suppo ort.

The Reactor R Tank is a 1/4 in nch thick qu uasi-hexag gonal 6061 plate aluminum consttruction nom minally 8 ft wide at co ore elevatioon increasin ng to 14 ft wide appro oximately 14 ft up from m the tank b bottom and d approxima ately 29 ft d deep.

A 3/4 thick Line er was insttalled interrior to the Reactor T Tank during g the opera ational histo ory atop su upport stripps/spacers.. The Bioshield conccrete suppo orting the tank varie es from 25 50 to 150 pcf with heavy 250 0 pcf concrrete around d core elevation. Th he BMRC also has moveable lead shield ds just ou utside the reactor ta ank at and d above ccore eleva ation.

Prelimminary acttivation analysis res ults for selected co omponents are prese ented below w.

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Plenuum Compo onents The Plenum, P an nd Plenum Chamber S Stand are located ben neath the aactive core and supporrted the corre while faccilitating co oolant flow. The Plenu um is cated from 3/8 thick fabric k 6061 al uminum a and interfacces with o other comp ponents withw stainle ess steel fasteners. Comp ponent iso otope abund dances and d scaling factors f are summarize ed in Table es 3-2 andd 3-3 beloww.

Tabble 3-2 Plenum m Chamberr Characteriz C zation Res sults Component C ID 03a Plenum P

Description Chhamber Waste W Weigght 54.82 lbs 2.49E+044 g Waste W Volumme 0.324 ft^3 9.16E+033 cc NRC Class A Fraactional Co-60 Fra actions Abuundance SF T1 T2 H-3 1.002E-04 1.08E-04 0.00 C-14 C 1.551E-03 1.61E-03 0.02 Fe-55 1.11E-02 1.18E-02 0.00 Co-60 C 9.440E-01 1.00E+00 0.11 Ni-59 3.552E-04 3.75E-04 0.00 Ni-63 4.74E-02 5.05E-02 0.11 Sr-90 S 1.999E-14 2.12E-14 0.00 Nb-94 3.886E-05 4.11E-05 0.15 Tc-99 T 2.335E-07 2.50E-07 0.00 Total T 1.0 00E+00 0.17 0.22 BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/2 2011 Pagge 11 of 28

Table e 3-3 Ple enum Cha amber Stan nd Chharacteriza ation Resullts Component ID 03c Plenum P

Description Chhamber Stand S

Waste Weig ght 31.93 lbs 1.45E+04 4 g Waste Volume 0.190 ft^3 5.37E+03 cc NRC C Class A Fractional Co-60 Fraactions Abuundance SF T1 T2 H-3 1.22E-11 1.28E-11 0.00 C-14 3.00E-03 3.15E-03 0.31 Fe-55 1.38E-02 1.45E-02 0.02 Co-60 9.52E-01 1.00E+00 1.12 Ni-59 2.49E-04 2.62E-04 0.01 Ni-63 3.08E-02 3.23E-02 0.72 Sr-90 1.09E-15 1.14E-15 0.00 Nb-94 2.25E-05 2.37E-05 0.93 Tc-99 5.77E-10 6.06E-10 0.00 Total 1.0 00E+00 1.24 1.86 BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/2 2011 Pag ge 12 of 28

Flappper The Flapper F is a hinged and a counte erbalanced plate that was shut d dung duringg forced co onvection and a open d during natuural circulaation. The inch thick componentt was fabric cated from 1100 grade e aluminum m with some e 304 stainless steel hinge h parts. Component isotope e abundancces and sccaling factorrs are summ marized in Table T 3-4 b below.

Tablee 3-4 Flap pper Chharacteriza ation Resullts Component ID 03b 0

Description Fla apper Waste Weig ght 19.00 l bs 8.62E+03 3 g Waste Volume 0.113 fft^3 3.19E+03 3 cc NRC C Class A Fracctional Co-60 Fraactions Abun ndance SF T1 T2 H-3 4.89 9E-12 5.00E-12 0.00 C-14 1.49 9E-03 1.53E-03 0.01 Fe-55 5.54 4E-03 5.67E-03 0.00 Co-60 9.77 7E-01 1.00E+00 0.10 Ni-59 1.10 0E-04 1.13E-04 0.00 Ni-63 1.54 4E-02 1.58E-02 0.03 Sr-90 1.25 5E-14 1.28E-14 0.00 Nb-94 3.84 4E-05 3.93E-05 0.14 Tc-99 5.52 2E-09 5.65E-09 0.00 Total 1.000E+00 0.15 0.13 BMRC Activation A Analysis A and d Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/2 2011 Pag ge 13 of 28

Ion Chamber C Guide G

The ioon chambe er guide is the t framew work of 2 incch, 6061 alluminum an ngles above e the Grid Plate boltted to the Plenum to o flange to locate the e ion cham mbers. Co omponent isotope ab bundances and scaling factorss are summ marized in Table T 3-5 below.

Tablee 3-5 Ion Chamb ber Guide Ch haracteriza ation Resullts Component C ID 04 Description Ion Chamber Guid de Waste W Weigght 7.21 5 lbs 3.27EE+03 g Waste W

Volume V 0.04 43 ft^3 1.21EE+03 cc NRRC Class A Frractional Co-600 Fractions Abbundance SF T1 T2 2

H-3 1.61E-11 1.70E-11 0.00 C-14 C 3.29E-03 3.46E-03 0.34 Fe-55 1.37E-02 1.45E-02 0.02 Co-60 C 9.49E-01 1.00E++00 1.12 Ni-59 2.72E-04 2.87E-04 0.01 Ni-63 3.32E-02 3.50E-02 0.78 Sr-90 S 1.30E-15 1.37E-15 0.00 Nb-94 2.47E-05 2.60E-05 1.02 Tc-99 T 6.33E-10 6.67E-10 0.00 Total T 1..00E+00 1.37 1.92 BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/22011 Pagge 14 of 28

Fission Chamb ber The Fission F Cha amber is a moveable componen nt located a at a core co orner and withdrawn w frrom the corre during po ower opera ation. The ffission cham mber is siz zed to fit inside its guide g and is fabricatted from 6 6061 alumin num.

Despite its relattively light weight, tra ansmutation n of Uranium to Plutonium is nott expected to drive wa aste class b because the e chamber was withdrawn duringg power operation.

o Componen nt isotope abundances and sccaling factorrs are summ marized in Table T 3-6 b below.

Table e 3-6 Fission C Chamber Chharacteriza ation Resullts Component C ID 05a Description Fissio on Chamberr Waste W Weigght 1..5 lbs 6.80EE+02 g Waste W

Volume V 0.00 09 ft^3 2.55EE+02 cc NRRC Class A Frractional Co-60 0 Fractions Abbundance SF T1 T2 2

H-3 7.25E-10 7.65E-10 0.00 C-14 C 3.59E-03 3.79E-03 2.06 Fe-55 1.49E-02 1.58E-02 0.10 Co-60 C 9.48E-01 1.00E+

+00 6.20 Ni-59 2.31E-04 2.44E-04 0.05 Ni-63 3.34E-02 3.52E-02 4.37 Sr-90 S 1.50E-14 1.58E-14 0.00 Nb-94 2.71E-05 2.86E-05 6.20 Tc-99 T 8.57E-10 9.04E-10 0.00 Total T 1..00E+00 8.31 10.667 3.2.2 Reac ctor Core Component C ts These e compone ents are subject s to the greate est neutron n flux and are conse equently the most activated componen nts. Thee reactor core comp ponents con nsist of the Grid Plate e, Grid Plate Plugs, Fiission Cham mber Guide e & Fission n Chamberr, Control E Elements, C Control Ele ement Shro ouds, and loower portioons of the Control C Elemment Drive S Shafts.

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Grid Plate The Grid G Plate is s a 5 inch thick t slab oof 1100 grad de aluminu um with a 6 by 6 array of 2.4 dia ameter holles milled tto accomm modate the lower fuel end fitting gs. 25 3/4 holes h are allso milled t o accommo odate the CControl Elem ment Shrou ud locating studs. 1/4 diameter d sttainless ste eel dowel pins are fixe ed on the upper u face e near each fuel ho ole to orie ent the fu uel assemb blies.

Comp ponent isottope abund dances and d scaling fa actors are summarize ed in Table e 3-7 beloww.

Tabble 3-7 Gridd Plate Characteriz C zation Res sults Component C ID 01 Description Grid Plate Waste W Weigght 88.336 lbs 4.01EE+04 g Waste W

Volume V 0.5222 ft^3 1.48EE+04 cc NRRC Class A Frractional Co-600 Fractions Abbundance SF T1 T2 2

H-3 4.71E-05 5.25E-05 0.00 C-14 C 3.44E-03 3.83E-03 2.04 Fe-55 2.06E-02 2.29E-02 0.14 Co-60 C 8.98E-01 1.00E++00 6.07 Ni-59 5.37E-04 5.98E-04 0.12 Ni-63 7.72E-02 8.60E-02 10.443 Sr-90 S 9.31E-14 1.04E-13 0.00 Nb-94 2.40E-05 2.67E-05 5.68 Tc-99 T 1.37E-08 1.53E-08 0.00 Total T 1..00E+00 7.83 16.664 BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/22011 Pagge 16 of 28

Grid Plate Plug gs and Wee ep Hole Plu ugs Grid Plate holes s not occup pied by fue el, Control Element S Shroud stud ds or experriments we ere filled with w Plugs. WMG conservativelyy assumed d the plugss are fab bricated frrom 6061 Aluminum. Comp ponent iso otope abund dances and d scaling fa actors are s ummarized d in Table 33-8 below.

Tab ble 3-8 Grid PPlate Plug Characteriz C zation Res sults Component C ID 02 Description Single Grid Plug Waste W Weigght 2.31 5 lbs 1.05EE+03 g Waste W

Volume V 0.01 4 ft^3 3.89EE+02 cc NRRC Class A Frractional Co-60 0 Fractions Abbundance SF T1 T2 2

H-3 7.36E-10 7.77E-10 0.00 C-14 C 3.72E-03 3.93E-03 2.00 Fe-55 1.55E-02 1.64E-02 0.10 Co-60 C 9.46E-01 1.00E+

+00 5.81 Ni-59 2.38E-04 2.52E-04 0.05 Ni-63 3.44E-02 3.63E-02 4.22 Sr-90 S 4.00E-15 4.23E-15 0.00 Nb-94 2.61E-05 2.76E-05 5.61 Tc-99 T 5.20E-10 5.49E-10 0.00 Total T 1..00E+00 7.65 10.113 BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/22011 Pagge 17 of 28

Fission Chamb ber Guide The Fission Ch hamber Gu uide is a 7 71/2 foot lon ng 3 inch diameter 6 6061 alumiinum tube with a fuell assembly lower end d fitting. It is located at a core corner and d is used too guide the e elevation n of the Fisssion Cham mber.

Comp ponent isottope abund dances and d scaling faactors are summarize ed in Tablee 3-9 beloww.

Tab ble 3-9 Fission Ch hamber Gu uide Characteriz C zation Res sults Component C ID 05a Description Fissio on Chamberr Guide Waste W Weigght 11.003 lbs 5.00EE+03 g Waste W

Volume V 0.0665 ft^3 1.85EE+03 cc NRRC Class A Frractional 0 Co-60 Fractions Abbundance SF T1 T2 2

H-3 7.25E-10 7.65E-10 0.00 C-14 C 3.59E-03 3.79E-03 2.06 Fe-55 1.49E-02 1.58E-02 0.10 Co-60 C 9.48E-01 1.00E+

+00 6.20 Ni-59 2.31E-04 2.44E-04 0.05 Ni-63 3.34E-02 3.52E-02 4.37 Sr-90 S 1.50E-14 1.58E-14 0.00 Nb-94 2.71E-05 2.86E-05 6.20 Tc-99 T 8.57E-10 9.04E-10 0.00 Total T 1..00E+00 8.31 10.667 BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/22011 Pagge 18 of 28

Control Elemen nts The neutron re eaction is controlled d by the elevation of the Co ontrol Elements. The 0.18Contro ol Elementss are fabriccated from a nickel-plated 80% silver - 15%  % indium - 5% cadm ium alloy. A flat alum minum transsition is useed to coup ple the platte-like Conttrol Elemen nt absorbeer with a 1 inch diame eter Drive Shaft. The e absorber is only 0.1 18 inches tthick (excluuding platin ng) by 4.85 5 inches wide by 29 inches tall. WMG asssumed the e five s attaching the transittion to the rivets e absorber are made from stain nless steel. Two sets s of Controll Elements exist with the set currrently store ed in the Reactor R Tank having a shorter exposure history tha an the firstt set.

Contrrol Elementt waste clas ss is most h heavily influ uenced by tthe activatioon of the th hin nickel cladding.

c Also, A cadmmium is con nsidered a toxic waste e, so the activated a Control C Ele ements sh hould be cconsidered mixed waste, partic cularly the first set of Control Ellements, w whose flat ttransitions wwere severred from the absorberr. Compon nent isotope e abundancces and sccaling factorrs are summ marized in Table T 3-10 below.

Tabble 3-10 Contro ol Element Characteriz C zation Res sults Component C ID 14a Sing gle Control Description Element E

Waste W Weigght 11.1550 lbs 5.06EE+03 g Waste W

Volume V 0.02 25 ft^3 7.10EE+02 cc NRRC Class A Frractional Co-60 0 Fractions Abbundance SF T1 T2 2

H-3 5.64E-05 5.95E-04 0.00 C-14 C 1.17E-04 1.24E-03 0.07 Fe-55 1.13E-02 1.20E-01 0.07 Co-60 C 9.47E-02 1.00E++00 0.62 Ni-59 7.05E-03 7.44E-02 1.47 Ni-63 8.87E-01 9.36E++00 116.48 Sr-90 S 1.11E-14 1.17E-13 0.00 Nb-94 8.89E-07 9.38E-06 0.20 Tc-99 T 5.93E-09 6.26E-08 0.00 Ag-108m A 1..46E+01 1.54E++02 Ag-108 A 1..30E+00 1.37E++01 Total T 1..00E+00 1.75 117.18 BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/2 2011 Pagge 19 of 28

Control Elemen nt Shrouds s Contrrol Elemen nt location is facilitatted by fra ame-like C Control Elem ment Shrouuds which act as a guide for the e absorberrs regardlesss of eleva ation.

The Control C Element Shrou uds have a rectangula ar cavity 5 inches wid de by 0.3055 inches thick and are e approxim mately 6 fee et tall. Thee locating sstuds and fasteners f are stainlesss steel while the rema ainder of the compone ent is assummed to be made from m 1100 gra ade alumin num. Com mponent iso otope abund dances and d scaling fa actors are s ummarized d in Table 3 3-11 below.

Tabble 3-11 Control C Ele ement Shrroud Characteriz C zation Res sults Component C ID 14b Control Element Description Shroud S

Waste W Weigght 87.1 2 lbs 3.95EE+04 g Waste W

Volume V 0.50 03 ft^3 1.42EE+04 cc NRRC Class A Frractional 0 Co-60 Fractions Abbundance SF T1 T2 2

H-3 1.48E-04 1.87E-04 0.00 C-14 C 2.80E-03 3.54E-03 2.17 Fe-55 3.60E-02 4.56E-02 0.32 Co-60 C 7.91E-01 1.00E+

+00 7.02 Ni-59 1.18E-03 1.49E-03 0.33 Ni-63 1.69E-01 2.14E-01 30.002 Sr-90 S 2.84E-13 3.59E-13 0.00 Nb-94 1.93E-05 2.44E-05 5.98 Tc-99 T 4.18E-08 5.28E-08 0.00 Total T 1..00E+00 8.49 37.336 BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/22011 Pag ge 20 of 28

Control Elemen nt Drive Shhafts Contrrol Elementt elevation is facilitate ed by the D Drive Shaftts. The Co ontrol Element Drive Shafts are e 1 inch d diameter 6 6061 alumiinum tube,, the lowerrmost part is activatted. The upper porrtion of th he shaft iss not activa ated. Com mponent is sotope abu ndances a and scaling factors for the lowerr shaft are summarized s d in Table 33-12 below.

Tab ble 3-12 Coontrol Elem ment Drive Shaft Characteriz C zation Res sults Component C ID 14c Control Element Description Drrive Shaft Waste W Weigght 1.1 7 lbs 5.30EE+02 g Waste W

Volume V 0.0007 ft^3 1.94EE+02 cc NRRC Class A Frractional Co-60 0 Fractions Abbundance SF T1 T2 2

H-3 7.36E-10 7.77E-10 0.00 C-14 C 3.72E-03 3.93E-03 2.00 Fe-55 1.55E-02 1.64E-02 0.10 Co-60 C 9.46E-01 1.00E+

+00 5.81 Ni-59 2.38E-04 2.52E-04 0.05 Ni-63 3.44E-02 3.63E-02 4.22 Sr-90 S 4.00E-15 4.23E-15 0.00 Nb-94 2.61E-05 2.76E-05 5.61 Tc-99 T 5.20E-10 5.49E-10 0.00 Total T 1..00E+00 7.66 10.113 BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/22011 Pagge 21 of 28

3.2.3 Expe erimental Facilities F

Special neutron n irradiationn facilities are provid ded for the e productio on of radioiisotopes an nd materials testing. Experimental facilitiess are locate ed in, arounnd and adja acent to thee active core. These experimen ntal facilitiess are assum med to be purged p of any a experim mental mate erials or isotope production materrial. Experimental facilities includ de the Pneu umatic Con nveyor, The ermal Colummn, Dry Ch hamber, Isootope Tube es, movable e Vertical S Standpipes,, and the Becquerel B Standpipe.

S The resullts for each h experime ental facilityy are discussed separrately below w.

Pneu umatic Con nveyors The Pneumatic c Conveyo ors are do ouble 6061 aluminum tube rrabbit assem mblies thatt could mo ove materia als from the hot cell to immediately adjaccent to the core in the horizontal plane usin ng differential air presssure.

Only the section ns closest to the core e are activaated. Com mponent iso otope abund dances and scaling factors f for the lower shaft are summarize ed in Tablee 3-13 below w.

Tabble 3-13 Pneumattic Convey yor Characteriz C zation Res sults Component C ID 07 Description Singlee Pneumaticc Conveyor Waste W Weigght 10.334 lbs 4.69EE+03 g Waste W

Volume V 0.0661 ft^3 1.74EE+03 cc NRRC Class A Frractional Co-60 0 Fractions Abbundance SF T1 T22 H-3 1.22E-11 1.28E-11 0.00 C-14 C 3.00E-03 3.15E-03 0.31 Fe-55 1.38E-02 1.45E-02 0.02 Co-60 C 9.52E-01 1.00E+

+00 1.12 Ni-59 2.49E-04 2.62E-04 0.01 Ni-63 3.08E-02 3.23E-02 0.72 Sr-90 S 1.09E-15 1.14E-15 0.00 Nb-94 2.25E-05 2.37E-05 0.93 Tc-99 T 5.77E-10 6.06E-10 0.00 Total T 1..00E+00 1.24 1.86 BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/2 2011 Pag ge 22 of 28

Thermmal Colum mn The Thermal Column C use ed a 6061 1 aluminum m exterior, graphite filled noseppiece imme ediately ad djacent to the active core to p provide the ermal neutrons to a large expe erimental sspace bene eath the h hot cell. 6 6061 alumiinum bails and rods s facilitatedd placeme ent of the reactor g grade graphhite blocks in the expe erimental sp pace. Onlyy the sectio on closest to o the core is substan ntially activ vated. Co omponent isotope ab bundances and scalin ng factors for the acttivated secction are su ummarized d in Table 3-14 beloww.

Tabble 3-14 Therma al Column Characteriz C zation Res sults Component C ID 08 Description Therrmal Column n Waste W Weigght 3162..0 lbs 1.43EE+06 g Waste W Volumme 21.9 98 ft^3 6.22EE+05 cc NRRC Class A Fractional F Co-60 0 Fractions Abbundance SF T1 T2 2

H-3 1.51E-07 1.60E-07 0.00 C-14 C 1.57E-03 1.66E-03 0.00 Fe-55 4.03E-02 4 4.26E-02 0.00 Co-60 C 9.46E-01 9 1.00E+

+00 0.00 Ni-59 1.07E-04 1.13E-04 0.00 Ni-63 1.25E-02 1.32E-02 0.00 Sr-90 S 0.00E+00 0.00E+

+00 0.00 Nb-94 8.22E-06 8 8.70E-06 0.00 Tc-99 T 4.34E-11 4 4.59E-11 0.00 Total T 1.00E+00 0.00 0.00 Dry Chamber C

Oppo osite the Thermal Column w was a sim milar but slightly la arger experrimental sp pace that was w not fille ed with graaphite. The e Dry Cham mber had a similar no osepiece, bu hield close tto the tank wall.

ut had a larrge lead sh Additional opera ational histtory informa ation is ne eeded to ch haracterize e Dry Cham mber compo onents otheer than the nosepiece.

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Isotoope Tubes, Vertical Standpipes S , and the BBecquerel Standpipe e Straigght and cu urved 6061 aluminum m tubes of different d diameters w were used for experim mental purp poses. So ome of the tubes loca ations could be chang ged relativee to the corre. Only the portions of the tube es closest too the core are activatted. Comp ponent isoto ope abunda ances and scaling factors for the activated sections are a summarrized in Tab bles 3-15 to o 3-17 below w.

Tabble 3-15 Isotoope Tube Characteriz C zation Res sults Component C ID 10 Description Iso otope Tube Waste W Weigght 1.8 81 lbs 8.20EE+02 g Waste W Volumme 0.01 1 ft^3 3.04EE+02 cc NRRC Class A Fractional F Co-60 0 Fractions Abundance SF T1 T2 2

H-3 7.29E-10 7 7.70E-10 0.00 C-14 C 3.69E-03 3 3.89E-03 2.01 Fe-55 1.53E-02 1 1.62E-02 0.10 Co-60 C 9.47E-01 9 1.00E+

+00 5.91 Ni-59 2.36E-04 2 2.50E-04 0.05 Ni-63 3.41E-02 3 3.60E-02 4.26 Sr-90 S 6.15E-15 6 6.50E-15 0.00 Nb-94 2.64E-05 2 2.78E-05 5.76 Tc-99 T 6.09E-10 6 6.43E-10 0.00 Total T 1.00E+00 1 7.82 10.227 BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/22011 Pag ge 24 of 28

Tab ble 3-16 Verticall Standpipe e Characteriz C zation Ressults Componen nt ID 11 Descriptionn Single Vertica al Standpipe Waste Weight 23 3.5 lbs 1.07EE+04 g Waste Volu ume 0.1 40 ft^3 3.96EE+03 cc N

NRC Class A Fractional Co-660 Fractions Abundance A SF F T T1 T T2 H-3 1.61E-11 1.70E E-11 0.000 C-14 3.29E-03 3.46E E-03 0.3 34 Fe-55 1.37E-02 1.45E E-02 0.002 Co-60 9.49E-01 1.00E+00 1.12 Ni-59 2.72E-04 2.87E E-04 0.0 01 Ni-63 3.32E-02 3.50E E-02 0.778 Sr-90 1.30E-15 1.37E E-15 0.000 Nb-94 2.47E-05 2.60E E-05 1.0 02 Tc-99 6.33E-10 6.67E E-10 0.0 00 Total 1.00E+00 1.3 37 1.992 BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/22011 Pag ge 25 of 28

Tabble 3-17 Isotoope Tube Characteriz C zation Res sults Component C ID 12 Bequerel B

Description Standpipe S

Waste W Weigght 33.0 08 lbs 1.50EE+04 g Waste W Volumme 0.19 96 ft^3 5.56EE+03 cc NRRC Class A Fractional F Co-60 0 Fractions Abundance SF T1 T2 2

H-3 1.61E-11 1 1.70E-11 0.00 C-14 C 3.29E-03 3 3.46E-03 0.34 Fe-55 1.37E-02 1 1.45E-02 0.02 Co-60 C 9.49E-01 9 1.00E+

+00 1.12 Ni-59 2.72E-04 2 2.87E-04 0.01 Ni-63 3.32E-02 3 3.50E-02 0.78 Sr-90 S 1.30E-15 1 1.37E-15 0.00 Nb-94 2.47E-05 2 2.60E-05 1.02 Tc-99 T 6.33E-10 6 6.67E-10 0.00 Total T 1.00E+00 1 1.37 1.92 BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/22011 Pag ge 26 of 28

4.0 COMPONE C ENT PACKA AGING REQUIREMEN NTS Detailedd radiation surveys must be performed p to finalize e the preliminary re esults presente ed in this analysis.

a Of O particula ar importancce is to ob btain surveyys of the h highly activatedd compone ents closestt to the active core. F February 20 011 compon nent surveyy use was ham mpered by detector saturation when w approoaching the e the highlyy activated core components. In addition, the e radiation surveys ide entify any hot spots on the rea actor components which h need to be taken n into conssideration during pacckaging off the components for transport and disposal.

The Con ntrol Eleme ent results and a discuss sion are likeely to drive e packaging g options fo or the BMRC. NRC Clas ss A mixed waste can be dispose ed at Clive, UT. How wever, base ed on results not n yet norrmalized to o measured d dose rate es, the Con ntrol Eleme ent waste cclass exceeds s that acce eptable at the Clive disposal site e. Even affter normalization to dose rates too be obtained in Ma ay 2011 of o a repre esentative Control Element, nu uclide concentration could d still exceeed NRC Class A limit s. Only the e Nevada T Test Site (N NTS) is curren ntly available for dispo osal of such h mixed wasste. The Plum Brook Reactor Fa acility had a ve ery similar predicamen nt that, afte er regulatorry effort, didd result in aacceptance e and disposall of non-De epartment of Energy mixed wasste. Key to o regulatoryy acceptancce at NTS is showing a nexus of o waste re esponsibilitty between n the BMR RC and fed deral governm ment entities s.

Preliminnary curie estimations e s indicate thet Grid P late, Contrrol Elementts, and Co ontrol Elementt Shrouds, may exceed the definiition of Low w Specific A Activity [II] (LSA II) matterial necessittating trans sport in an NRC N licens sed Type B cask. A ssingle 8-120 0 cask linerr has sufficien nt volume to t accomm modate all the t highly activated componentts with min nimal segmentation. The shielding g afforded by b the 8-12 20 is also ssufficient fo or all the h highly activatedd compone ents. The la arge payloa ad capacityy of and 8-120 minimiizes handlin ng of components, reduc cing the timme and do ose necesssary to pacckage the ccomponents for disposall. Tubing sections s andd other itemms exceedi ng 5 feet lo ong need to o be segme ented to accom mmodate th he internal height of the disposa al liner andd furthermo ore, to increase packagin ng efficienccy of the ac ctivated com mponents. A packaging efficienccy of up to 15%

(waste volume v / liner internal volume)) can be a achieved w with segme entation of long components and ca areful packaging. .

The minimally activated componentsc s of the reactor sstructure, periphery and experimental facilitties can be e packaged for dispo osal as Lo ow Specificc Activity (L LSA) material. Disposal liner/conta ainer and transportat t ion packag ge selection n will be based upon the e detailed radiation r su urveys obta ained during g decommisssioning. A Activation oof the reactor tank and other o minim mally activa ated compo onents mayy have loca alized hotspots.

Howeve er, the activ vation levels s in these components c s are low eenough succh that loca alized hotspotss will not change th he overall approach to remova al and disposal of these components, nor approach an ny regulatorry limits.

BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/22011 Pag ge 27 of 28

5.0 REFERENC R CES

1. ORNL L CCC-254 4, ANISN-O ORNL One-D Dimensiona al Discrete Ordinates Transsport Code with Anisottropic Scatttering, Oakk Ridge Nattional Laborratory, April 1991.

2. DLC-185, BUGL LE-96, Coupled, 47-Ne eutron, 20 G Gamma-Ra ay Cross Sectioon Library derived d from m ENDF/B--VI for LWR R Shielding and Pressure Vesse el Dosimetrry Applications, Oak R Ridge National Laborattory Radiattion Shieldding Inform mation Cente er, July 19999.

3. ORNL L/CCC-371, ORIGEN 2.2, Isotop pe Generation and Dep pletion Cod de, Matrixx Exponenttial Method, Oak Ridge e National laboratory, May 2002..

4. Evans, J.C., et al.,

a NUREG G 3474, Lon ng-Lived Acctivation Pro oducts in Reac ctor Materials, Pacific Northwest N LLaboratorie es, August 11984.

5. NURE EG/BR-020 04, Instructions for Commpleting NRCs Uniform Low-Levvel Radiooactive Was ste Manifesst, July 199 98.

BMRC Activation A Analysis A andd Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/2 2011 Pagge 28 of 28

APPENDIX A ORRIGEN Matterial Comp positions BMRC Activation A Analysis A Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/2 2011 Page P A1

3 Stainless Steel (N 304 NUREG 347 74) g lbs x Density = 8.0 03 3

, 5001 3 cm ft Aveerage NURE EG 3474 30 04 Stainles ss Steel Co omposition n Weigght Weight RSSICC Fracttion RSICC Fraction Element Number E (g/g g) Elemennt Number (g/g)

Li 30 0000 1.30E E-07 Sr 380000 2.00E-07 N 70 0000 4.52E E-04 Y 390000 5.00E-06 Na 110000 9.70E E-06 Zr 400000 1.00E-05 Al 130000 1.00E E-04 Nb 410000 8.90E-05 Cl 170000 7.00E E-05 Mo 420000 2.60E-03 K 190000 3.00E E-06 Ag 470000 2.00E-06 Ca 200000 1.90E E-05 Sb 510000 1.23E-05 Sc 210000 3.00E E-08 Cs 550000 3.00E-07 Ti 220000 6.00E E-04 Ba 560000 5.00E-04 V 230000 4.56E E-04 La 570000 2.00E-07 Cr 240000 1.84E E-01 Ce 580000 3.71E-04 Mn 250000 1.53E E-02 Sm 620000 1.00E-07 Fe 260000 7.06E E-01 Eu 630000 2.00E-08 Co 270000 1.41E E-03 Tb 650000 4.70E-07 Ni 280000 1.00E E-01 Dy 660000 1.00E-06 Cu 290000 3.08E E-03 Ho 670000 1.00E-06 Zn 300000 4.57E E-04 Yb 700000 2.00E-06 Ga 310000 1.29E E-04 Lu 710000 8.00E-07 As 330000 1.94E E-04 Hf 720000 2.00E-06 Se 340000 3.50E E-05 W 740000 1.86E-04 Br 350000 2.00E E-06 Pb 820000 6.70E-05 Rb 370000 1.00E E-05 Th 900000 1.00E-06 U 920000 2.00E-06 BMRC Activation A Analysis A Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/2 2011 Page P A2

C Carbon Steeel (NUREG G 3474) g lbs x Density = 7.8 87 3

, 4991 3 cm ft A

Average NUUREG 3474 4 Carbon Steel S Compposition Weigght Weight RSSICC Fracttion RSICC Fraction E

Element Number (g/g g) Elemennt Number (g/g)

Li 30 0000 3.00E E-07 Sr 380000 1.50E-07 N 70 0000 8.40E E-05 Y 390000 2.00E-05 C* 80 0000 2.90E E-03 Zr 400000 1.00E-05 Na 110000 2.30E E-05 Nb 410000 1.88E-05 Al 130000 3.30E E-04 Mo 420000 5.60E-07 Cl 170000 4.00E E-05 Ag 470000 2.00E-06 K 190000 1.20E E-05 Sb 510000 1.10E-05 Ca 200000 1.40E E-05 Cs 550000 2.00E-07 Sc 210000 2.60E E-07 Ba 560000 2.73E-04 Ti 220000 2.00E E-06 La 570000 1.00E-07 V 230000 8.00E E-05 Ce 580000 1.00E-06 Cr 240000 1.70E E-03 Sm 620000 1.70E-08 Mn 250000 1.02E E-02 Eu 630000 3.10E-08 Fe 260000 9.80E E-01 Tb 650000 4.50E-07 Co 270000 1.22E E-04 Ho 670000 8.00E-07 Ni 280000 6.60E E-03 Yb 700000 1.00E-06 Cu 290000 1.27E E-03 Lu 710000 2.00E-07 Zn 300000 1.00E E-04 Hf 720000 2.10E-07 Ga 310000 8.00E E-05 Ta 730000 1.30E-07 As 330000 5.32E E-04 W 740000 5.50E-06 Se 340000 7.00E E-07 Pb 820000 8.20E-04 Br 350000 8.50E E-07 Th 900000 1.80E-07 Rb 370000 4.80E E-05 U 920000 2.00E-07

• Thhe carbon weight perrcent is tak ken from Page P 2, Grrade 65 [Grade 4550], for typical pressurre vessel stteel becaus se there is no n carbon liisted in NUREG 34 474.

BMRC Activation A Analysis A Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/22011 Page P A3

B Bioshield Cooncrete (NU UREG 3474 4) g lbs x Density = 2.3 30 3

, 1444 3 cm ft Aveerage NURE EG 3474 Bioshield Co oncrete Coompositionn Weig ght Weight RSSICC Fracttion RSICC Fraction Element Number E (g/g g) Elemennt Number (g/g)

H 10 0000 6.10EE-3 Rb 370000 3.50E-5 Li 30 0000 2.00EE-5 Sr 380000 4.38E-4 B 50 0000 2.00EE-5 Y 390000 1.82E-5 N 70 0000 1.20EE-4 Zr 400000 7.10E-5 O* 80 0000 5.02EE-1 Nb 410000 4.30E-6 Na 110000 7.39E E-3 Mo 420000 1.03E-5 Mg* 120000 2.46E E-3 Pd 460000 3.00E-6 Al 130000 3.10E E-2 Ag 470000 2.00E-7 Si 140000 1.68E E-1 Cd 480000 3.00E-7 P 150000 5.00E E-3 Sn 500000 7.00E-6 S 6000 16 3.10EE-3 Sb 510000 1.80E-6 Cl 170000 4.50E E-5 Cs 550000 1.30E-6 K 190000 7.50E E-3 Ba 560000 9.50E-4 Ca 200000 1.83E E-1 La 570000 1.30E-5 Sc 210000 6.50E E-6 Ce 580000 2.43E-5 Ti 220000 2.12E E-3 Sm 620000 2.00E-6 V 230000 1.03E E-4 Eu 630000 5.55E-7 Cr 240000 1.09E E-4 Tb 650000 4.10E-7 Mn 250000 3.77E E-4 Dy 660000 2.30E-6 Fe 260000 3.90E E-2 Ho 670000 9.00E-7 Co 270000 9.80E E-6 Yb 700000 1.40E-6 Ni 280000 3.80E E-5 Lu 710000 2.70E-7 Cu 290000 2.50E E-5 Hf 720000 2.20E-6 Zn 300000 7.50E E-5 Ta 730000 4.40E-7 Ga 310000 8.80E E-6 W 740000 1.40E-6 As 330000 7.90E E-6 Pb 820000 6.10E-5 Se 340000 9.20EE-7 Th 900000 3.50E-6 Br 350000 2.40EE-6 U 920000 2.70E-5

• Thee oxygen an nd magnesium weight percents are a calculate ed since the ese nucclides are exxcluded fro om NUREG 3474.

BMRC Activation A Analysis A Reporrt 10-102D--RE-138 Reev. A Compon nent Charac cterization 05/2 2011 Page P A4

A Aluminum 6 6061 g lbs x Density = 2.77 3

, 1699 3 cm ft Avera age Alumin num 6061 Compositio C on Weigght Weight RSSICC Fraction RSICC Fraction Element Nummber (g/g g) Elemennt Number (g/g)

B 50 0000 8.43E E-7 Fe 260000 7.00E-3 N 70 0000 1.37E E-4 Co 270000 1.72E-4 Mg 120 0000 1.00E E-2 Ni 280000 4.04E-4 Al 130 0000 9.66E E-1 Cu 290000 2.75E-3 Si 140 0000 6.00E E-3 Zn 300000 2.50E-3 Ti 220 0000 1.50E E-3 Nb 410000 3.40E-5 Cr 240 0000 1.95E E-3 Mo 420000 5.60E-7 Mn 250 0000 1.50E E-3 G

Graphite g lbs x Density = 1.77 3

, 1066 3 cm ft Avverage Gra aphite Com mposition Weigght Weight RSSICC Fraction RSICC Fraction Element Nummber (g/g g) Elemennt Number (g/g)

H 10 0000 3.89E E-5 Mg 120000 1.00E-4 B 50 0000 5.00E E-7 Cl 170000 2.50E-5 C 60 0000 9.89E E-1 Ti 220000 1.00E-3 O 80 0000 1.29E E-3 Fe 260000 8.00E-3 F 90 0000 2.50E E-5 Co 270000 5.00E-5 Na 110 0000 1.00E E-4 Cu 290000 1.00E-4 BMRC Activation A Analysis A Reporrt 10-102D--RE-138 Re ev. A Compon nent Charac cterization 05/2 2011 Page P A5

BMRC Decommissioning Plan Revision 0 APPENDIX B - SITE CHARACTERIZATION REPORT

Site Characterization Report Buffalo Materials Research Center Prepared for:

Buffalo Material Research Center Office of Environment, Health, and Safety Services Completed by:

4490 Old William Penn Hwy Murrysville, PA 15668 July 30, 2011

BMRC Characterization Report Revision 0 Summary of Changes Revisions to the BMRC Characterization Report will be tracked when revisions are issued.

Changed sections will be identified by special demarcation in the margin. A summary description of each revision will be noted in the following table.

Revision Date Description of Change Number 0 July 30, 2011 Initial Issue iv

BMRC Characterization Report Revision 0 Table of Contents ACRONYMS AND ABBREVIATIONS ....................................................................................... ii

1. EXECUTIVE

SUMMARY

.................................................................................................1

2. SCOPE AND BACKGROUND ..........................................................................................3 2.1 SCOPE .................................................................................................................................. 3 2.2 FACILITY BACKGROUND ..................................................................................................... 3

3.

SUMMARY

OF TECHNICAL APPROACH.....................................................................9 3.1 RADIONUCLIDES OF CONCERN............................................................................................ 9 3.2 RELEASE CRITERIA ........................................................................................................... 12 3.2.1 Soils....................................................................................................................... 12 3.2.2 Surfaces ................................................................................................................. 13 3.3 SURVEY DESIGN................................................................................................................ 14 3.4 SURVEY PACKAGE IDENTIFICATION ................................................................................. 15 3.5 SURVEY INSTRUMENTATION ............................................................................................. 16 3.6 MINIMUM DETECTABLE ACTIVITY ................................................................................... 16 3.7 PRELIMINARY SURVEY UNITS AND CLASSIFICATIONS ..................................................... 17

4. CURRENT RADIOLOGICAL CONDITIONS ................................................................21 4.1 ADMINISTRATIVE WING.................................................................................................... 21 4.1.1 Control Deck ......................................................................................................... 21 4.1.2 Gamma Deck......................................................................................................... 22 4.2 SUB-BASEMENT ................................................................................................................ 23 4.3 CONTAINMENT WING........................................................................................................ 23 4.3.1 Control Deck ......................................................................................................... 23 4.3.2 Gamma Deck......................................................................................................... 25 4.3.3 Neutron Deck ........................................................................................................ 26 4.4 CONCRETE CORES ............................................................................................................. 26 4.4.1 Floor Cores............................................................................................................ 27 4.4.2 Bioshield Cores ..................................................................................................... 30 4.5 EXTERIOR SOILS AND BUILDING SURFACES ..................................................................... 33 4.5.1 Sub-Surface Borings ............................................................................................. 33 4.5.2 Surface Water and Groundwater Samples............................................................. 34 4.5.3 Gamma Radiation Scan......................................................................................... 38 4.5.4 BMRC Rooftops.................................................................................................... 39 4.5.5 Tritium and Carbon-14 Smears ............................................................................. 39 4.6 TANK FARM ...................................................................................................................... 40 4.7 VENTILATION .................................................................................................................... 43 4.8 MISCELLANEOUS SAMPLE AREAS .................................................................................... 43 4.8.1 N16 Tank Vault..................................................................................................... 43 4.8.2 1K Tank................................................................................................................. 45

5. HAZARDOUS MATERIALS ...........................................................................................46

6. QUALITY ASSURANCE AND QUALITY CONTROL.................................................47 6.1 GENERAL PROVISIONS ...................................................................................................... 47 6.1.1 Written Procedures................................................................................................ 47 6.1.2 Instrumentation Selection, Calibration, and Use................................................... 47 6.1.3 Chain of Custody................................................................................................... 48 iv

BMRC Characterization Report Revision 0 6.1.4 Independent Review of Survey Results................................................................. 48 6.2 TRAINING .......................................................................................................................... 48 6.3 SAMPLE ANALYSIS ............................................................................................................ 48

7. LIMITATIONS..................................................................................................................49 7.1 HIGH BACKGROUND AREAS ............................................................................................. 49 7.2 INTERNAL ACTIVITY ......................................................................................................... 49

8. REFERENCES ..................................................................................................................50 List of Tables Table 3-1 Radionuclides of Concern .............................................................................................. 9 Table 3-2 Site-Specific Radionuclides of Concern at BMRC ...................................................... 11 Table 3-3 NRC Surface Soil Screening Levels............................................................................. 12 Table 3-4 NRC Regulatory Guide 1.86 Levels for Surfaces ........................................................ 13 Table 3-5 NRC Screening Levels for Building Surfaces.............................................................. 14 Table 3-6 Survey Instrumentation ................................................................................................ 16 Table 3-7 Instrumentation MDA .................................................................................................. 17 Table 3-8 Survey Unit Summary .................................................................................................. 19 Table 4-1 Administrative Wing Control Deck Survey Results Summary.................................... 22 Table 4-2 Administrative Wing Gamma Deck Survey Results Summary ................................... 22 Table 4-3 Sub-Basement Survey Results Summary ..................................................................... 23 Table 4-4 Containment Wing Control Deck Survey Results Summary ....................................... 24 Table 4-5 Containment Wing Gamma Deck Survey Results Summary....................................... 25 Table 4-6 Containment Wing Neutron Deck Survey Results Summary ...................................... 26 Table 4-7 Concrete Floor Cores Analytical Results Summary..................................................... 29 Table 4-8 Bioshield Cores Analytical Results Summary ............................................................. 32 Table 4-9 Soil Borings Analytical Results Summary ................................................................... 35 Table 4-10 Tank Farm Analytical Sample Results Summary ...................................................... 42 Table 4-11 Ventilation Systems Survey Results Summary .......................................................... 43 Table 4-12 N16 Tank Vault Analytical Results Summary ........................................................... 44 Table 4-13 1K Tank Analytical Results Summary ....................................................................... 45 iv

BMRC Characterization Report Revision 0 List of Figures Figure 1: University of Buffalo South Campus .............................................................................. 4 Figure 2: Control Deck Layout ....................................................................................................... 6 Figure 3: Gamma Deck Layout....................................................................................................... 7 Figure 4: Neutron Deck Layout ...................................................................................................... 8 Figure 5: Floor Core Locations..................................................................................................... 28 Figure 6: Bioshield Core Locations .............................................................................................. 30 Figure 7: Soil Bore Locations ....................................................................................................... 33 Figure 8: Gamma Walkover.......................................................................................................... 38 Figure 9: Tank Farm Layout ......................................................................................................... 41 Figure 10: N16 Tank Vault Sample Locations ............................................................................. 44 List of Appendices Appendix A - BMRC Characterization Photographs iv

BMRC Characterization Report Revision 0 List of Acronyms and Abbreviations AEC Atomic Energy Commission ACM Asbestos Containing Material BF Bioshield Face BMRC Buffalo Materials Research Center D&D Decontamination and Dismantlement DCGL Derived Concentration Guideline Level DP Decommissioning Plan DPM Disintegrations per Minute DQO Data Quality Objectives FSS Final Status Survey HSA Historical Site Assessment MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MDA Minimum Detectable Activity MDC Minimum Detectable Concentration MWt Megawatt Thermal NaI Sodium Iodide NIST National Institute for Standards and Testing NRC Nuclear Regulatory Commission NUREG Nuclear Regulatory Guide PULSTAR Pulse Training and Reactor R-77 Reactor License Number 77 RFP Request For Proposal RSO Radiation Safety Officer SSCs Systems Structures Components SUNY State University of New York UB University of Buffalo v

BMRC Characterization Report Revision 0

1. EXECUTIVE

SUMMARY

This report describes the current radiological status of the facility to support the development of the Buffalo Materials Research Center (BMRC) Decommissioning Plan (DP). The sections that follow describe the scope, background, and technical approach used during the characterization of the BMRC. The BMRC contains a 2-Megawatt Thermal (MWt) pool-type research reactor licensed by the United States Nuclear Regulatory Commission (NRC) under reactor license number 77 (R-77) and is located on the South Campus of the State University of New York (SUNY) at Buffalo, NY.

The data collected during this characterization will be utilized to support the development of a DP in accordance with the following NRC regulatory guidance documents: 1) NUREG-1537, Part 1, Chapter 17, Guidelines for Preparing and Reviewing Applications for Licensing of Non-Power Reactors; 2) NUREG-1757, Consolidated NMSS Decommissioning Guidance; and 3)

NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM).

The data will specifically assist in the development of radiation safety protocols and project dose estimates; estimation of expected radiological waste volumes; methodologies for removal and packaging of radiological waste; and disposal options for radioactively impacted materials generated during decommissioning activities conducted under NRC License R-77. This data will also be incorporated into the decision making process for the development of Final Status Survey (FSS) Data Quality Objectives (DQOs) described in MARSSIM. The FSS provides the closeout documentation required by the NRC in order to terminate the BMRC Reactor License.

For the characterization effort, ENERCON developed a Characterization Plan to establish the measurement and sampling technical approach along with the project quality assurance requirements. This plan was reviewed and approved by the BMRC Operating Committee.

Radiological work permits were issued by the Universitys Radiation Safety Officer (RSO) for all survey areas, as necessary. The BMRC facility was separated into survey units with survey instructions for each unit. The survey instructions were implemented by a qualified Health Physics technician and the data was documented on a radiological survey. This report only contains a summary of the survey results. All completed survey packages are stored at the BMRC for future reference.

1

BMRC Characterization Report Revision 0 Radiological measurements indicate that levels of residual materials on the building surfaces are less than release criteria established by the NRC in Regulatory Guide 1.86. Soil samples indicate that remediation of soils will not be required outside of the facility; however, radionuclides above the release criteria were detected in the N16 Vault which does not have a concrete floor covering the bedrock. Residual radioactive materials were only found in areas indicated by the Historical Site Assessment. The structures and components that contain the majority of the residual radioactive materials are the activated reactor components and bioshield, the former and current liquid waste tanks, and the reactor water cooling system. Residual radioactive materials were also identified in the Hot Cell and laboratory fume hoods.

2

BMRC Characterization Report Revision 0

2. SCOPE AND BACKGROUND 2.1 Scope The BMRC characterization scope was to determine the nature and extent of residual radioactive materials and was accomplished through the following methods:

x Radiological measurements on building surfaces and structures x Concrete cores on the reactor bioshield to determine concrete activation x Exterior sub-surface soil borings x Interior concrete floor cores x Liquid samples from legacy and current waste holding tanks The activation analysis of reactor components and pool liner is not included in this characterization report. The Component Activation Analysis scope of work was contracted by the University to WMG, Inc. of Fishkill, NY. The Component Activation Analysis Report, along with this characterization report, will be appended to the BMRC DP for approval by the NRC.

2.2 Facility Background The UB is licensed by NRC to possess radioactive materials within the BMRC. The BMRC reactor is located on the South Campus of the University at Buffalo (UB) in the city of Buffalo, New York, as shown in Figure 1.

The reactor was defueled in 2005 and is beginning the Decontamination and Dismantlement (D&D) process to decommission the facility for eventual termination of the NRC license. This Characterization Report is in support of a DP developed in accordance with Chapter 17 of the Nuclear Regulatory Commission (NUREG -1537) Part 1 Guidance for Preparing and Reviewing Applications for Licensing of Non-Power Reactors (NRC 1996).

This report describes the purpose, scope and technical approach used for the characterization of the BMRC. The purpose of the characterization is to assess the current radiological status of the facility for the development of the Decommissioning Plan. This data is also incorporated into the decision-making process for development of FSS DQOs described in MARSSIM.

3

BMRC Characterization Report Revision 0 Figure 1: University of Buffalo South Campus The BMRC building is comprised of an administrative/laboratory wing (administrative wing) and a vapor containment building (containment building) with three levels in each. The three levels are: neutron deck (lowest level, i.e. sub-grade), gamma deck (2nd level, i.e. middle level),

and control deck (top floor). The neutron deck of the administrative wing (a.k.a. sub-basement) houses numerous systems, structures, and components (SSCs) that support operation of the reactor. The remainder of the administrative wing contains 20 rooms which are a mixture of offices, laboratories, utility rooms, and classrooms. See Figures 2, 3, and 4 at the end of this section.

The facility was constructed in 1959, went into operation on March 24, 1960, and was operated by The Western New York Nuclear Research Center, Inc, a subsidiary of the State University of New York, under Atomic Energy Commission (AEC) license number R-77. The facility 4

BMRC Characterization Report Revision 0 licensee was changed to the Nuclear Science and Technology Center followed by a change to the Buffalo Materials Research Center in 1973 and 1985, respectively. The licensed power was one MWt without forced convection cooling and up to two MWt possible for short times. In 1991, the reactor was retrofitted with a 0.75 inch thick aluminum liner. This new liner supplements the original 0.25 inch thick aluminum liner and there is a 0.75 inch air gap between the two liners.

The reactor core is located in a pool-type tank that is 29 feet deep with horizontal dimensions of 13 feet by 7 feet, located in the containment building of the BMRC. The tank contains approximately 13,700 gallons of demineralized water. The reactor fuel consisted of standard zircalloy clad Pulse Training and Reactor (PULSTAR) type fuel elements.

Several events have occurred at the BMRC during its operational history that has led to the determination that the buildings, structures, and immediate surrounding soils are impacted from historical operations at BMRC. The following events are documented in the BMRC Historical Site Assessment (HSA).

x In 1975, and again in 1989, a leak developed in the primary cooling piping at the point where the pipe penetrates the reactor pool tank. The leakage is thought to have traversed the primary piping exterior to a point where it emerged from the foundation concrete in Room N16 and then to the sumps associated with the Cooling Water Equipment Room (Room N03).

x In 1993, the Primary Coolant Heat Exchanger located in Room N02 of the Laboratory Wing lower level, developed a primary to secondary leak that permitted small amounts of coolant to transfer to the water cascading in the cooling tower. The cooling tower and its above grade concrete structure have since been removed. The concrete basin was left in place and backfilled to grade.

x In the Laboratory Wing, the drains in the Vertical Accelerator Room (Room 115) are posted as contaminated as a result of drains backing up during operation.

x Events were discussed during interviews that relate to airborne contamination events within the Containment Building. These events resulted in areas or levels of the Containment Building being designated in whole or part as contaminated. One event was an airborne release of Co60 to the entire Neutron Deck. Decontamination occurred in all areas of the Containment Building except for areas near the ceiling and the ceiling. The second event was a release of Sr90 that contaminated areas of Room 202. The walls adjacent to the California hood in this room have plexiglass mounted to provide shielding from the strontium. There was also one event where H3 was released in the Control Deck Upper Airlock and Upper Fuel Vault.

5

BMRC Characterization Report Revision 0 Figure 2: Control Deck Layout 6

BMRC Characterization Report Revision 0 Figure 3: Gamma Deck Layout 7

BMRC Characterization Report Revision 0 Figure 4: Neutron Deck Layout 8

BMRC Characterization Report Revision 0

3.

SUMMARY

OF TECHNICAL APPROACH Radiological characterization surveys were performed to establish the current radiological status of the BMRC. The site characterization survey included radiological surface measurements on building surfaces and equipment in addition to volumetric measurements on surrounding soils, floors, and the reactor bioshield. The characterization surveys were designed, performed, evaluated, and documented in accordance with the following sections.

The technical approach for the characterization of the building surfaces of the BMRC was developed to meet MARSSIM criteria for FSS. A minimum quantity of survey locations for each survey unit was defined to meet a 95% certainty of detection using nonparametric statistical analyses contained in MARSSIM. Survey locations within the building were selected in Visual Sample Plan Version 6.0 using either a random or systematic triangular grid placement method.

Survey technicians were also instructed to use professional judgment in identifying additional biased locations (i.e., fume hoods, elevated measurements, lab benches) to assure that the nature and extent of any residual radioactive materials present were adequately defined.

The technical approach for exterior soils was developed using MARSSIM guidance to determine the impacts, if any, to surface water and groundwater.

3.1 Radionuclides of Concern Based upon facility operating history and historical radiological analyses, both alpha and beta/gamma emitting radionuclides may be present in the BMRC. Evaluation of radionuclides of concern identified in NUREG/CR-3474, NUREG/CR-4289, and those detected via laboratory analyses from samples collected at the facility yields a substantial list of 60 radionuclides. This list is presented in Table 3-1.

Table 3-1 Radionuclides of Concern Half Life Detected at Nuclide (yr) Citation Source BMRC Ac-228 7.00E-04 N/A Yes Ag-108m 4.18E+02 NUREG/CR-3474 Yes Ag-110m 6.84E-01 NUREG/CR-3474 No Am-241 4.32E+02 NUREG/CR-4289 No Ar-39 2.69E+02 NUREG/CR-3474 No 9

BMRC Characterization Report Revision 0 Half Life Detected at Nuclide (yr) Citation Source BMRC Ba-133 1.05E+01 NUREG/CR-3474 No C-14 5.73E+03 NUREG/CR-3474 Yes Ca-41 1.03E+05 NUREG/CR-3474 No Ce-141 8.90E-02 NUREG/CR-3474 No Cl-36 3.01E+05 NUREG/CR-3474 No Cm-244 1.81E+01 NUREG/CR-4289 No Co-57 7.44E-01 N/A Yes Co-58 1.94E-01 NUREG/CR-3474 Yes NUREG/CR-3474 Co-60 5.27E+00 Yes NUREG/CR-4289 Cr-51 7.58E-02 NUREG/CR-3474 No Cs-134 2.06E+00 NUREG/CR-3474 No Cs135 2.30E+06 NUREG/CR-3474 No NUREG/CR-3474 Cs-137 3.02E+01 Yes NUREG/CR-4289 NUREG/CR-3474 Eu-152 1.36E+01 Yes NUREG/CR-4289 NUREG/CR-3474 Eu-154 8.59E+00 Yes NUREG/CR-4289 Eu-155 4.96E+00 NUREG/CR-3474 No Fe-55 2.70E+00 NUREG/CR-3474 No Fe-59 1.22E-01 NUREG/CR-3474 No NUREG/CR-3474 H-3 1.23E+01 Yes NUREG/CR-4289 Hf-178m 3.00E+01 NUREG/CR-3474 No Ho-166m 1.20E+03 NUREG/CR-3474 No NUREG/CR-3474 I-129 1.57E+07 No NUREG/CR-4289 Kr-81 2.10E+05 NUREG/CR-3474 No Kr-85 1.07E+01 NUREG/CR-3474 No Mn-53 3.70E+06 NUREG/CR-3474 No Mn-54 8.56E-01 NUREG/CR-3474 No Mo-93 3.50E+00 NUREG/CR-3474 No Nb-92m 2.78E-02 NUREG/CR-3474 No NUREG/CR-3474 Nb-94 2.03E+04 No NUREG/CR-4289 NUREG/CR-3474 Ni-59 7.50E+04 No NUREG/CR-4289 NUREG/CR-3474 Ni-63 1.00E+02 Yes NUREG/CR-4289 Np-237 2.14E+06 NUREG/CR-4289 No Pb-205 1.51E+07 NUREG/CR-3474 No Pm-145 1.77E+01 NUREG/CR-3474 No Pu-238 8.78E+01 NUREG/CR-4289 Yes NUREG/CR-3474 Pu-239 2.41E+04 Yes NUREG/CR-4289 Pu-240 6.60E+03 NUREG/CR-4289 Yes Ra-226 1.60E+03 N/A Yes Sb-124 1.65E-01 NUREG/CR-3474 No 10

BMRC Characterization Report Revision 0 Half Life Detected at Nuclide (yr) Citation Source BMRC Sc-46 2.29E-01 NUREG/CR-3474 No Se-79 6.50E+04 NUREG/CR-3474 No Sm-146 1.00E+08 NUREG/CR-3474 No Sm-151 9.30E+01 NUREG/CR-3474 No Sn-121m 5.00E+00 NUREG/CR-3474 No Sr-89 1.38E-01 N/A Yes NUREG/CR-3474 Sr-90 2.88E+01 Yes NUREG/CR-4289 Tb-158 1.50E+02 NUREG/CR-3474 No NUREG/CR-3474 Tc-99 2.13E+05 No NUREG/CR-4289 Th-228 1.91E+00 N/A Yes Th-232 1.41E+10 N/A Yes U-233 1.59E+05 NUREG/CR-3474 No U-234 2.46E+05 N/A Yes U-238 4.47E+9 N/A Yes Zn-65 6.69E-01 NUREG/CR-3474 No Zr-93 1.53E+06 NUREG/CR-3474 No Thirty-six (36) of the 60 listed radionuclides have not been detected in samples collected from the facility. Removing those radionuclides which have not been detected; radionuclides with a half-life less than one (1) year; and, naturally occurring radionuclides which have been detected only at background concentrations pares this list down to a site-specific list of fourteen (14) radionuclides as presented in Table 3-2.

Table 3-2 Site-Specific Radionuclides of Concern at BMRC Radionuclide Half Life (yr) Emission Area(s) of Interest Ag-108m 4.18E+02  Soil; SSCs Am-241 4.32E+02  Tank sediment C-14 5.73E+03 Laboratory areas Co-60 5.27E+00  Soil; SSCs; Bioshield Cs-137 3.02E+01 * Soil; SSCs; Bioshield Eu-152 1.36E+01  Soil; SSCs; Bioshield Eu-154 8.59E+00  Soil; SSCs; Bioshield H-3 1.23E+01 Soil; SSCs, Bioshield Ni-63 1.00E+02 Soil; SSCs; Bioshield Pu-238 8.78E+01  Tank sediment Pu-239 2.41E+04  Tank sediment Pu-240 6.60E+03  Tank sediment Ra-226 1.60E+03  Discrete sources Sr-90 2.88E+01 SSCs; Ventilation systems; Soil emission from Ba-137m progeny 11

BMRC Characterization Report Revision 0 3.2 Release Criteria The BMRC characterization effort included evaluation of SSCs and of soils surrounding the facility. While release criteria are not required for the characterization process, consideration of likely criteria in selecting field instrumentation and determining action levels is important. Two sets of criteria were considered; one set for surface soils, and one for SSCs.

3.2.1 Soils NRC License Termination Screening Levels contained in Appendix H of NUREG-1757 Volume 2 for surface soils were considered during the characterization process at BMRC. These criteria, listed in Table 3-3, provide a point of comparison for the soil data collected around the BMRC.

In addition to the NRC Screening Value, a second screening value column has been added to compare the NRC values at 25 mrem/year to the State of New Yorks 10 mrem/year dose release criteria established in Radioactive Materials Guidance Document, DSHM-RAD-05-01, Cleanup Guidelines for Soils Contaminated with Radioactive Materials, issued by the Department of Environmental Conservation. Of the site-specific radionuclides of interest, surface soil screening levels are available for all but Ag-108m. Additional evaluation of criteria for this radionuclide will be required in the DP for the FSS.

Table 3-3 NRC Surface Soil Screening Levels NRC Screening Value Screening Value for Surface Soils Scaled for NY State*

Radionuclide (pCi/g) (pCi/g)

Ag-108m None None Am-241 None None C-14 12 4.8 Co-60 3.8 1.52 Cs-137 11 4.4 Eu-152 8.7 3.48 Eu-154 8 3.2 H-3 110 44 Ni-63 2,100 840 Pu-238 2.5 1 Pu-239 2.3 0.92 Pu-240 None None Ra-226 0.7 0.28 Sr-90 1.7 0.68 New York State dose requirement is 10 mrem/yr (40% of the NRC Values) 12

BMRC Characterization Report Revision 0 3.2.2 Surfaces For the BMRC characterization effort, unrestricted release of material and equipment prior to a final status survey considers criteria from NRC Regulatory Guide 1.86. These criteria were considered in calculating the instrumentation counting sensitivities in Section 3.1.6. The Regulatory Guide 1.86 criteria are more conservative than the NRC screening levels for building surfaces presented in Appendix H of NUREG-1757 Volume 2 and are listed in Table 3-4. Final Status Survey criteria, if needed, will use those criteria in NUREG-1757 presented in Table 3-5.

Table 3-4 NRC Regulatory Guide 1.86 Levels for Surfaces Averageb c Maximumb d Removableb e Radionuclidea dpm/100 cm2 dpm/100 cm2 dpm/100 cm2 U-nat, U-235, U-238 and associated decay products 5,000 15,000 1,000 Transuranics, Ra-226, Ra-228, Th-230, Th-228, Pa-231, 100 300 20 Ac-227, I-125, I-129 Th-nat, Th-232, Sr-90, Ra-223, Ra-224, U-232, I-126, I-1,000 3,000 200 131, I-133 Beta-gamma emitters (nuclides with decay modes other than alpha emission or spontaneous fission) except Sr-90 5,000 15,000 1,000 and others noted above.

a Where surface contamination by both alpha and beta-gamma emitting nuclides exists, the limits established for alpha- and beta-gamma-emitting nuclides should apply independently.

b As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detector for background, efficiency, and geometric factors associated with the instrumentation.

c Measurements of average contaminant should not be averaged over more than 1 square meter. For objects of less surface area, the average should be derived for each object.

d The maximum contamination level applies to an area of not more than 100 cm2.

e The amount of removable radioactive material per 100 cm2 of surface area should be determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency. When removable contamination on objects of less surface area is determined, the pertinent levels should be reduced proportionally and the entire surface.

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BMRC Characterization Report Revision 0 Table 3-5 NRC Screening Levels for Building Surfaces NRC Screening Value for Building Surfaces Radionuclide (DPM/100cm2)

Ag-108m None Am-241 None C-14 3,700,000 Co-60 7,100 Cs-137 28,000 Eu-152 None Eu-154 None H-3 120,000,000 Ni-63 1,800,000 Pu-238 None Pu-239 None Pu-240 None Ra-226 None Sr-90 8,700 Consideration of the release criteria for building surfaces is important when selecting remediation strategies and for developing the final status survey. The release criteria for material and equipment are also important when estimating the quantity of material and equipment that will be disposed as radioactive waste.

3.3 Survey Design Characterization surveys were designed to determine the presence of residual radioactive materials at the BMRC. In each area, professional judgment was used in conjunction with a systematic or random survey design for selection of locations that may have increased potential for residual radioactive materials. These locations were then surveyed for total alpha, total beta, removable alpha, and removable beta and general area gamma radiation dose rates.

In the survey design, interior surfaces of the administrative wing were initially considered a minimum of MARSSIM Class 3. The entire containment wing was considered MARSSIM Class1 with the exception of the containment ceiling, which was considered Class 2. At a minimum, 29 survey locations were identified per survey package. Survey packages could consist of more than one room and a survey unit could contain more than one survey package.

Additional survey locations were sometimes identified by the survey technician during the 14

BMRC Characterization Report Revision 0 survey for collection of biased survey data. As such, a minimum of 29 survey locations were identified per survey package.

The identified survey locations within a survey package received a static total alpha/beta measurement, an alpha/beta removable wipe sample, and a beta/gamma scan of the 1 square meter area surrounding the location. Gamma dose rate measurements were performed 1-meter above identified floor locations.

Background measurements were collected in air within the selected room, or alternately, during the source check in the established technician staging area (Room 210/212).

The status of these rooms at the time of the survey is documented in the photographs in Appendix A.

3.4 Survey Package Identification During the investigation, a consistent identification system was used to ensure both the uniqueness and clarity in evaluation location and an identification of data collected. This section describes the protocol that was employed in naming the evaluation locations. Each location was assigned a unique identifier that included:

x Class (1, 2, or 3) x Building Level o C for the control deck level o G for the gamma deck level o N for the neutron deck level x Survey unit number - sequential number relative to the class and building level x Room identifier as historically identified This resulted in a unique identifier for each survey package. For example, the survey package 2C1-215 represents Room 215 as being survey Class 2 on the control deck in survey unit 1. In this case, adjacent Room 213 is also part of survey unit 1 and is included in the survey package 2C1-213.

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BMRC Characterization Report Revision 0 3.5 Survey Instrumentation Radiological survey instrumentation was selected to ensure that sensitivities were sufficient to detect the expected radionuclides at the minimum detection requirements. The instrumentation was calibrated in accordance with approved procedures to National Institue of Science and Technology (NIST) traceable standards. The calibration was checked daily using Th-230 and Tc-99 for alpha and beta/gamma measurements, respectively. A list of the survey instrumentation, along with the type of radiation detected, calibration sources, and the instrument use, is provided in Table 3-6.

Table 3-6 Survey Instrumentation Radiation Calibration Instrument Detector Type Detected Source Use Ludlum Model 43-68 Surface Static Measurements; Ludlum Model 2221 Gas Proportional Beta/Gama Tc-99 Beta/Gamma scan (126 cm2 area) measurements Ludlum Model 43-89 ZnS Ludlum Model 2360 coated Plastic Scintillator Alpha/Beta Th-230/Tc-99 Alpha static measurements (126 cm2 area)

Ludlum Model 19 Internal NaI Gamma Cs-137 General area exposure rates Ludlum Model 43-10-1 Ludlum Model 3030E Alpha/Beta Th-230/Tc-99 Swipe/smear counting ZnS internal detector 3.6 Minimum Detectable Activity The minimum detectable activity (MDA), also referred to minimum detectable concentration (MDC) is the concentration of radioactivity that an instrument can be expected to detect at a 95 percent confidence level. For instruments performing direct measurements and for laboratory analyses, the MDA goal was 10-50% of applicable release criteria. For static (direct) surface measurements, with conventional detectors, the MDA was calculated using the formula:

3  3.29 Rb Ts 1  Ts / Tb º MDA (dpm/100cm2) = << >>

<<¬ Ts Ht >>1/4 Variables:

Rb = Background count rate (cpm)

Tb = Background count time (min)

Ts = Sample Count Time (min)

Ht = Total Instrument Efficiency (MARSSIM section 6.6.1) 16

BMRC Characterization Report Revision 0 The data used to calculate the MDA for the instrumentation used during the BMRC characterization are shown on the survey records in stored at the BMRC. The typical a priori MDA is listed in Table 3-7.

Table 3-7 Instrumentation MDA Average MDA Instrument Detector Type Radiation Detected (dpm/100 cm2)

Alpha 47 Ludlum Model 2360 Ludlum Model 43-89 Beta 612 Alpha 59 Ludlum Model 2360 Ludlum Model 43-68 Beta 393 Alpha 12 Ludlum Model 3030E Ludlum Model 43-10-1 Beta 159 Ludlum Model 2221 Ludlum Model 43-68 Beta 271 3.7 Preliminary Survey Units and Classifications The BMRC was segmented into logical areas of similar history and/or construction as survey units. Utilizing guidance available in the MARSSIM, an initial classification of either impacted or non-impacted was assigned to each survey unit. Non-impacted areas are areas where historical information and/or process knowledge does not indicate the existence of residual radioactive material, and does not indicate any reason to suspect the presence of residual radioactive materials. Impacted areas are areas where historical information and/or process knowledge indicates a potential presence of residual radioactive materials. During the characterization phase of the BMRC, all areas within the building were considered as impacted.

Based on the levels of potential residual radioactive materials, impacted areas are further divided into Class 1, Class 2 or Class 3 designations. Class 1 areas have the greatest potential for residual activity while Class 3 areas have the least potential for impacted areas. Each classification will typically be bounded by areas classified one step lower to provide a buffer zone around the higher class. Exceptions occur when an area is surrounded by a significant physical barrier that would make transport of residual activity unlikely from one area to the adjacent area. In such cases, each area is classified solely on its own merit using the most reliable information available. The class definitions provided below are from Section 4.4 of the MARSSIM.

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BMRC Characterization Report Revision 0 Class 1 Areas that have, or had prior to remediation, a potential for radioactive contamination (based on site operating history) or known contamination (based on previous radiological surveys). Examples of Class 1 areas include: 1) site areas previously subjected to remedial actions, 2) locations where leaks or spills are known to have occurred, 3) former burial or disposal sites, 4) waste storage sites, and 5) areas with contaminants in discrete solid pieces of material high specific activity. Note that areas containing contamination in excess of the DCGLw prior to remediation should be classified as Class 1 areas.

Class 2 These areas have, or had prior to remediation, a potential for radioactive contamination or known contamination, but are not expected to exceed the DCGLw. To justify changing an areas classification from Class 1 to Class 2, the existing data (from the HSA, scoping surveys, or characterization surveys) should provide a high degree of confidence that no individual measurement would exceed the DCGLw. Other justifications for this change in an areas classification may be appropriate based on the outcome of the DQO process. Examples of areas that might be classified as Class 2 for the final status survey include: 1) locations where radioactive materials were present in an unsealed form (e.g., process facilities), 2) potentially contaminated transport routes, 3) areas downwind from stack release points, 4) upper walls and ceilings of some buildings or rooms subjected to airborne radioactivity, 5) areas where low concentrations of radioactive materials were handled, and 6) areas on the perimeter of former contamination control areas.

Class 3 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, based on site operating history and previous radiological surveys.

Examples of areas that might be classified as Class 3 include buffer zones around Class 1 or Class 2 areas, and areas with very low potential for residual contamination but insufficient information to justify a non-impacted classification.

The preliminary list of survey units is presented in Table 3-8.

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BMRC Characterization Report Revision 0 Table 3-8 Survey Unit Summary Survey Class Deck Unit Room ID Room Description 1 C 1 200 Electronics shop 1 C 2 201NW Control Deck Northwest quadrant 1 C 3 202 Laboratory 1 C 4 203 Laboratory 1 C 5 204 Fan Room 1 C 6 205 Reactor Control 1 C 7 206 Office 1 C 8 207 Superintendent Office 1 C 9 201A Fuel Storage Vault 1 C 9 201B Airlock 1 C 10 201NE Control Deck Northeast quadrant 1 C 11 201SE Control Deck Southeast quadrant 1 C 12 201SW Control Deck Southwest quadrant 1 C 12 S242 Stairs from Gamma Level to Control Level 1 C 13 201HC Hot Cell Roof 1 C 14 201U Area above airlock 1 C 15 203U Area above rooms 202/203 1 C 16 204U Area above room 204 1 C 17 205U Area above room 205 1 C 18 206U Area above room 206 1 C 19 207U Area above room 207 2 C 1 213 Laboratory 2 C 1 215 Laboratory 2 C 2 Ceiling Containment ceiling 3 C 1 208 General Managers Office 3 C 1 209 Mens Locker and Change Area 3 C 1 210 Conference Room 3 C 1 212 Conference Room(cont) 3 C 1 214 Secretary Office 3 C 1 216 Directors Office 3 C 1 221 Womens Rest Area 3 C 1 208A Reception Area 3 C 1 208B General Manager Closet 3 C 1 209A Janitors Closet 3 C 1 209B Mens Toilet Room 3 C 1 214A Secretary Office Closet 3 C 1 216A Directors Office Closet 3 C 1 221A Womens Toilet Area 3 C 1 C241 Corridor 3 C 1 OTB Vertical Accelerator upper 3 C 1 S240 Stairs Hall 1 G 1 101 Airlock 1 G 1 101A Janitors Closet 1 G 2 102NW Gamma Deck 1 G 2 S123 Stairs 19

BMRC Characterization Report Revision 0 Survey Class Deck Unit Room ID Room Description 1 G 3 103 Chemistry Lab 1 G 4 104 Medical Lab 1 G 5 105 Work Room 1 G 6 106 Change Room 1 G 7 107 Hot Cell Service Room 1 G 8 108 Hot Chemistry Lab 1 G 9 109 Storage Area 1 G 10 113 Shop 1 G 11 115 Laboratory 1 G 12 114 Health Physics/Counting Room 1 G 13 115A Vertical Accelerator lower 1 G 14 102NE Gamma Deck NE area 1 G 15 102E Gamma Deck East area 1 G 16 102S Gamma Deck South area 1 G 17 Hotcell Hot Cell 1 G 18 102W Gamma Deck West area 2 G 1 117 Classroom 2 G 1 118 Horizontal Accelerator 3 G 1 110 Electrical Room 3 G 1 111 Air Conditioning Equipment Room 3 G 1 111A Janitors Closet 3 G 1 S121, S122, & C119 Stair Hall 1 N 1 N01 -A Neutron Deck 1 N 2 N16 N16 tank room 1 N 3 SN07 Stairs 1 N 4 N02AIR Air shaft to exterior 1 N 5 N03 Cooling Water Equipment Room 1 N 6 N03PIT 1000 gal tank pit 1 N 7 N02-North Air Conditioning Equipment Room - North section 1 N 8 N02-South Air Conditioning Equipment Room - South section 1 N 9 N01-B Neutron Deck 1 N 10 N01-C Neutron Deck 1 N 11 N01-D Neutron Deck 1 N 12 N01-E Neutron Deck 1 N 13 N01-F Neutron Deck 1 N 14 N01-G Neutron Deck 2 N 1 10K 10,000 gal Tank Room 3 N 1 N04 Air Conditioning Equipment Room 3 N 1 N05 Air Plenum 3 N 1 SN06 Stair Hall 20

BMRC Characterization Report Revision 0

4. CURRENT RADIOLOGICAL CONDITIONS The following sections summarize the results of the radiological characterization effort. Per MARSSIM guidance all measurement results, whether positive, negative, or less than MDA, were included in the calculations of the mean, maximum, and standard deviation. The results are summarized by survey unit. The completed survey packages have been retained for future reference, but not included in this report.

4.1 Administrative Wing Several rooms in the administrative wing have radioactive materials stored in them. Due to elevated background from the radioactive material, it was difficult to determine the presence of residual radioactive materials. It is ENERCONs belief that these areas do not have residual radioactive materials present and after the stored radioactive material is packaged and shipped off site, these measurements will be verified.

The rooms were surveyed for total alpha, total beta, removable alpha, and removable beta.

Twenty-nine (29) building surface locations were selected in each room. In addition, biased locations were selected (e.g. fume hoods, drains, and sinks) at the discretion of the survey technician.

Each subsection below contains a table which summarizes characterization survey results for that area.

4.1.1 Control Deck The following rooms on the control deck of the administrative wing show no elevated measurements of residual radioactive materials:

x Locker Rooms (Rooms 209, 221 and 221A) x Conference Room (Room 212) x Offices (Rooms 208, 210, 214 and 216) x Reception Area (Room 208A) x Mens Toilet Room (Room 209B) x Janitor Closet (Room 209A) x Two labs (Rooms 213 and 215) 21

BMRC Characterization Report Revision 0 Table 4-1 Administrative Wing Control Deck Survey Results Summary Removable Alpha Removable Beta Static Alpha Static Beta Measurement Measurement Results Measurement Measurement Results(DPM/100cm2)

(DPM/100cm2) Results(DPM/100cm2) Results(DPM/100cm2)

Survey Std Std Std Std Unit # Mean Max Dev Mean Max Dev Mean Max Dev Mean Max Dev 2C1 172 1 15 3 11 101 23 19 67 16 218 1537 341 3C1 441 0 6 1 8 90 17 17 139 28 258 4310 673 4.1.2 Gamma Deck The following rooms on the gamma deck of the administrative/laboratory wing show no elevated measurements of residual radioactive materials:

x Electrical Service Room (Room 110) x Mechanical Room (Room 111) x Machine Shop (Room 113)

Several rooms on the gamma deck of the administrative/laboratory wing contain radioactive materials. These rooms are:

x Horizontal Accelerator (Room 118) with an adjacent cave area x Class Room (Room 114) x Health Physicist Office (Room 117) x Vertical Accelerator Room (Rooms 115 and 115A)

Table 4-2 Administrative Wing Gamma Deck Survey Results Summary Removable Alpha Removable Beta Static Alpha Static Beta Measurement Results Measurement Results Measurement Results Measurement Results (DPM/100cm2) (DPM/100cm2) (DPM/100cm2) (DPM/100cm2)

Survey Std Std Std Std Unit # Mean Max Dev Mean Max Dev Mean Max Dev Mean Max Dev 1G10 76 1 6 1 4 64 11 27 89 17 89 659 175 1G11 76 1 12 2 2 21 5 37 67 19 200 979 210 1G12 78 0 3 1 2 37 6 23 84 17 619 4436 1072 1G13 76 0 3 1 8 59 15 31 75 19 1019 17159 3007 2G1 144 0 6 1 12 96 24 18 50 13 242 960 272 3G1 174 0 6 1 3 48 10 27 58 14 100 1330 217 22

BMRC Characterization Report Revision 0 4.2 Sub-Basement The sub-basement, located below the gamma deck of the administrative/laboratory wing includes rooms that house the Holdup Tank Cubicle (Room N16), the Cooling Water Equipment Room and the Facility Sumps (Room N03), the lower Air Conditioning Equipment Room (Room N02),

the Air Handling Duct Room (Room N04) and access to the 10K tank enclosure (Room N05).

Room NO3 houses a 1000 gallon tank. This tank contains approximately 6 inches of sludge with elevated levels of radioactive materials. Room N16 houses a 5,000 gallon N16 decay tank which is internally contaminated with radioactive materials.

Table 4-3 Sub-Basement Survey Results Summary Removable Alpha Removable Beta Static Alpha Static Beta Measurement Measurement Results Measurement Results Measurement Results Results (DPM/100cm2)

(DPM/100cm2) (DPM/100cm2) (DPM/100cm2)

Survey Std Std Std Std

1. Mean Max Mean Max Mean Max Mean Max Unit Dev Dev Dev Dev 1N2 29 1 6 2 47 112 27 6 25 8 4664 6941 1076 1N4 33 1 6 2 2 32 7 27 67 18 2477 49163 8823 1N5 38 1 3 1 5 48 12 27 58 17 3569 16147 3511 1N6 27 1 3 2 13 213 41 410 823 99 24499 364513 74158 1N7 33 1 6 2 16 324 58 24 75 20 5472 55469 12542 1N8 34 1 3 1 1 21 5 8 50 12 241 1056 272 2N1 30 0 0 0 43 149 35 21 50 15 No Data No Data No Data 3N1 93 1 6 2 3 64 11 7 69 14 299 3463 577 4.3 Containment Wing The containment wing of the BMRC is accessible via airlocks from the administrative wing control and gamma deck levels. Exterior access is via a truck door on the gamma deck.

4.3.1 Control Deck The Control Deck (top level) houses a dual-hook over head crane with one 10-ton and one 2-ton hook on tracks at the ceiling level. Crane access to the lower levels is via a floor plate on the Control Deck and a webbing cover on a floor penetration from the Gamma Deck to the Neutron Deck. In addition to the reactor pool, the Control Deck houses:

x Reactor Control Room (Room 205) x Superintendents Office (Room 207) 23

BMRC Characterization Report Revision 0 x Two labs (Rooms 202 and 203) x Fan Room (Room 204) x Office (Room 206) x Airlock (Room 201B) for access to the upper or First Floor of the Administrative Wing x New Fuel Storage Room (Room 201A) located within the doors of the airlock Most of these rooms were being used to store legacy radioactive wastes when surveys were performed. The elevated background from the radioactive material storage areas made it difficult to determine the presence of residual radioactive materials on building surfaces in these areas. The limited radiological measurements collected in these areas did not identify removable or fixed radiological activity; therefore, the building surfaces are not believed to be contaminated. Following the removal of these legacy materials for offsite disposal, additional measurements will be collected in these areas. Room 202 was the location of a release of Sr-90 that contaminated the walls adjacent to the California hood. These walls are affixed with Plexiglas coverings as a mitigating measure to provide beta shielding.

Table 4-4 Containment Wing Control Deck Survey Results Summary Removable Alpha Removable Beta Static Alpha Static Beta Measurement Measurement Results Measurement Results Measurement Results Results (DPM/100cm2)

(DPM/100cm2) (DPM/100cm2) (DPM/100cm2)

Survey Std Std Std Std Unit # Mean Max Dev Mean Max Dev Mean Max Dev Mean Max Dev 1C1 30 1 3 1 7 64 15 126 238 60 231 2084 508 1C2 40 1 3 1 8 74 16 80 179 37 200 704 241 1C3 41 2 9 3 41 138 36 64 149 39 12226 241518 42125 1C4 38 1 12 3 21 96 23 41 139 34 83 2001 355 1C5 38 1 6 2 18 74 21 122 218 54 225 649 196 1C6 38 1 6 2 23 96 24 109 248 46 148 1470 265 1C7 38 2 12 3 28 170 36 101 188 37 476 13016 2094 1C8 45 1 6 1 6 48 12 84 218 55 220 704 212 1C9 29 4 18 5 14 80 20 531 1280 320 10904 71732 17243 1C10 34 1 12 3 10 96 21 111 208 45 470 2692 477 1C11 38 2 9 2 11 59 17 95 407 95 127 697 177 1C12 68 4 18 4 6 59 12 111 575 81 1733 32484 4619 1C14 26 1 9 2 4 37 10 172 267 38 1101 2050 362 1C15 28 1 6 2 3 37 10 55 409 79 393 1270 331 1C16 29 1 6 2 5 32 11 72 226 62 1028 3122 821 1C17 29 3 12 3 3 53 12 149 936 209 733 3267 828 1C18 29 4 15 4 11 69 21 181 259 48 849 1667 377 1C19 25 0 3 1 4 32 9 211 326 50 832 1177 200 24

BMRC Characterization Report Revision 0 4.3.2 Gamma Deck The middle level or Gamma Deck houses:

x Hot Cell x Hot Cell Work Room (Room105) x Locker Room (Room 106) x Hot Chemical Lab (Room 108) x Medium Chemical Lab (Room 104) x Activation Analysis Lab (Room109) x Lower Fuel Vault (Room 101A) x Chemical Lab (Room 103) x Airlock (Room 101) for access between the administrative wing and containment This level also has a Truck Lock that provides vehicular and equipment access to the outside.

Prior to 1979, this level housed the waste compaction system. Wastes generated at BMRC and under the site-wide radiological program were collected and compacted for volume reduction in this system.

Most rooms on the Gamma Deck were being used to store radioactive material which makes it difficult to determine the presence of residual radioactive materials on building surfaces Additional measurements will be collected on building surfaces in these areas following waste removal to verify residual radioactivity levels. All items inside the Hot Cell, including the Hot Cell, should be considered radioactive waste.

Table 4-5 Containment Wing Gamma Deck Survey Results Summary Removable Alpha Removable Beta Static Alpha Static Beta Measurement Results Measurement Results Measurement Results Measurement Results (DPM/100cm2) (DPM/100cm2) (DPM/100cm2) (DPM/100cm2)

Survey Std Std Std Std Unit # Mean Max Dev Mean Max Dev Mean Max Dev Mean Max Dev 1G1 44 2 9 3 2 21 5 96 208 42 188 407 125 1G2 30 2 9 2 2 59 11 62 134 33 694 3074 1068 1G3 38 2 6 3 7 64 16 50 169 35 204 2947 518 1G4 38 1 6 2 6 43 13 7 60 15 4851 33706 6862 1G5 38 3 15 4 3 43 9 124 192 34 324 2011 407 1G6 40 2 12 3 7 64 14 155 259 45 1802 3849 942 1G7 38 2 6 2 11 69 20 56 209 51 1124 7763 1632 1G8 39 3 15 4 20 165 39 91 217 44 6727 32540 8022 1G9 42 2 12 3 2 27 6 93 167 37 10227 151217 25043 25

BMRC Characterization Report Revision 0 4.3.3 Neutron Deck The Neutron Deck (lower level) is divided by caging into two sections. The open area includes access to the Patient Treatment Room which is lined with wood. The Patient Treatment Room was included in the original design of the BMRC to be a location to provide neutron treatments to cancer patients. There have been no indications that it was ever used for this purpose. The enclosed area includes the rerouted Primary Coolant Piping and access to the seven Beam Storage Tubes. The Beam Storage Tubes extend 10 feet through the Containment Building wall into the surrounding soil. They were used as storage-for-decay of high dose reactor components or experiments. At the point where they exit the containment building wall, they are approximately 10 feet underground. The balance of the neutron deck is open area. The Neutron Deck is the area where the majority of radioactive waste materials awaiting disposal were being stored. No removable radioactive material was found during the characterization survey. Upon removal of the stored radioactive materials, a more detailed investigation will be performed.

Table 4-6 Containment Wing Neutron Deck Survey Results Summary Removable Alpha Removable Beta Static Alpha Static Beta Measurement Measurement Results Measurement Results Measurement Results Results (DPM/100cm2)

(DPM/100cm2) (DPM/100cm2) (DPM/100cm2)

Survey Std Std Std Std Unit # Mean Max Dev Mean Max Dev Mean Max Dev Mean Max Dev 1N1 19 1 6 2 12 48 18 139 334 87 2092 14488 3389 1N9 35 2 9 3 8 53 14 336 593 89 11993 167287 35497 1N10 38 0 6 1 2 21 6 0 0 0 21152 200489 45808 1N11 13 0 0 0 1 5 2 1 10 3 192605 1888420 512482 1N12 41 5 15 4 18 122 29 86 198 53 181 1801 429 1N13 38 1 6 2 7 48 13 258 443 79 3216 22262 4683 1N14 38 1 6 2 17 122 27 109 308 68 674 3202 962 4.4 Concrete Cores Concrete core samples were collected to assess volumetric activity in the neutron deck floor and the bioshield. In addition, the concrete cores in the neutron deck were used to determine if a preferential pathway existed between the neutron deck floor and the bedrock. The concrete cores were field scanned for gross gamma radioactivity prior to being cut into approximate 1 inch thick 26

BMRC Characterization Report Revision 0 cross-sections for radioanalysis. Cross-sections that exhibited elevated count rates were sent offsite for analysis. If no elevated count rate was detected at any cross-sectional location in the core, at a minimum, the end cross-section was submitted.

4.4.1 Floor Cores Six (6) locations on the Neutron Deck floor were chosen to collect 3-inch diameter concrete cores. The locations are shown in Figure 5. Field scans of the cores indicated no residual radioactive materials were present. Floor core #1 was collected through the proposed elevator pit and was therefore not as thick as the remaining floor cores. Locations #5 through #6 were cored through the concrete floor to the bedrock. During construction, the concrete floor of the neutron deck was poured directly on the excavated bedrock to an approximate depth of 30 inches. The concrete cores proved that the concrete was poured directly on the bedrock because no sand or gravel bedding layer was found in any location. Therefore a preferential pathway for radioactive materials does not exist between the concrete floor and the bedrock. The analytical results indicate essentially background activities for the evaluated radionuclides. A summary of the laboratory analytical results for the floor cores is included in Table 4-7.

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BMRC Characterization Report Revision 0 Figure 5: Floor Core Locations 28

BMRC Characterization Report Revision 0 Table 4-7 Concrete Floor Cores Analytical Results Summary Core ID> FLOOR 1 FLOOR 2 FLOOR 3 FLOOR 4 FLOOR 5 FLOOR 6 Max Max Max Max Max Max Max Max Max Max Max Max Activity MDC Activity MDC Activity MDC Activity MDC Activity MDC Activity MDC Radionuclide (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g)

Ag-110M -0.03 0.04 0.01 0.06 0.00 0.02 -0.01 0.02 -0.02 0.05 -0.01 0.05 Co-58 -0.04 0.06 -0.04 0.06 -0.01 0.02 0.00 0.03 0.00 0.06 -0.02 0.06 Co-60 0.09 0.06 0.00 0.07 0.01 0.03 0.00 0.02 -0.01 0.05 0.00 0.05 Cs-134 -0.03 0.05 0.01 0.06 0.00 0.03 0.00 0.03 0.02 0.05 0.00 0.05 Cs-137 0.01 0.06 0.00 0.06 0.00 0.02 0.00 0.03 0.01 0.06 0.00 0.06 Eu-152 0.08 0.15 -0.03 0.13 0.00 0.07 0.03 0.09 0.05 0.14 -0.08 0.15 Eu-154 0.02 0.09 -0.02 0.08 0.01 0.03 0.01 0.04 0.02 0.08 -0.03 0.09 H-3 3.12 0.53 3.64 0.51 2.57 0.51 2.08 0.53 5.55 0.57 2.37 0.54 K-40 5.99 0.25 6.09 0.53 3.47 0.28 3.65 0.20 4.46 0.45 2.81 0.55 La-140 -0.12 0.12 -0.04 0.20 -0.06 0.07 -0.01 0.11 -0.01 0.18 -0.01 0.20 Mn-54 0.04 0.06 -0.01 0.05 0.01 0.03 0.00 0.03 -0.02 0.05 -0.02 0.06 Ra-226 1.66 1.02 2.02 0.43 2.02 1.02 Sb-124 0.00 0.06 -0.01 0.07 0.00 0.03 0.02 0.04 0.00 0.07 -0.02 0.07 Th-228 0.28 0.11 0.22 0.08 0.12 0.04 0.14 0.05 0.21 0.08 0.17 0.09 Th-232 0.37 0.17 0.29 0.19 0.20 0.08 Note: Blank field indicates no reported result 29

BMRC Characterization Report Revision 0 4.4.2 Bioshield Cores Six (6) locations along two (2) walls of the bioshield face (BF) on the neutron deck were chosen for core samples as shown in Figure 6. Three (3) cores were advanced in-line with the beam tubes until contact was made with the original 1/4 inch aluminum liner at the inner surface of the bioshield. This resulted in a total length for each core of approximately 70 inches.

Evaluation of the data for the bioshield core samples indicates measureable activity in certain areas exceeds 12 inches in depth. The charts below show the Co-60 analytical results of the two sets of bioshield core data in a graphical manner of depth vs. activity. The Co-60 results are used as being representative of the overall results. Chart 1 presents Co-60 results vs. depth from the liner for the bioshield core samples on wall 4 and Chart 2 presents the data for bioshield core samples on wall 5. A summary of the bioshield core analytical results is presented in Table 4-8.

Figure 6: Bioshield Core Locations 30

BMRC Characterization Report Revision 0 Chart 1: Bioshield Cores Wall 4 Chart 2: Bioshield Cores Wall 5 31

BMRC Characterization Report Revision 0 Table 4-8 Bioshield Cores Analytical Results Summary Core ID> BF1-1 BF1-2 BF1-3 BF2-1 BF2-2 BF2-3 Max Max Max Max Max Max Max Max Max Max Max Max Activity MDC Activity MDC Activity MDC Activity MDC Activity MDC Activity MDC Radionuclide (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g)

Ag-110M 0.02 0.24 0.08 0.28 0.06 0.16 0.00 0.05 0.03 0.18 -0.01 0.04 Co-58 0.10 0.35 0.23 0.42 0.01 0.24 0.03 0.08 0.06 0.27 0.00 0.06 Co-60 104.90 0.30 175.90 0.32 50.08 0.18 6.68 0.07 69.10 0.22 4.41 0.05 Cs-134 0.03 0.21 0.02 0.25 0.06 0.15 0.00 0.05 0.05 0.16 0.01 0.04 Cs-137 0.10 0.25 0.19 0.30 0.05 0.17 0.01 0.06 0.06 0.20 0.01 0.05 Eu-152 5.01 0.53 10.75 0.63 3.89 0.37 0.31 0.13 4.50 0.42 0.29 0.11 Eu-154 2.95 0.35 3.85 0.39 1.78 0.25 0.17 0.08 1.94 0.28 0.05 0.06 H-3 141.00 2.37 290.00 3.70 154.00 2.47 8.01 1.18 110.00 2.03 19.70 0.97 K-40 0.61 0.29 1.09 0.35 0.70 0.21 La-140 0.11 0.41 0.06 0.56 0.08 0.33 -0.02 0.13 0.07 0.34 0.01 0.10 Mn-54 0.12 0.31 0.17 0.36 0.01 0.21 0.02 0.07 0.04 0.24 0.00 0.05 Ra-226 Sb-124 0.11 0.27 0.08 0.33 0.07 0.19 0.00 0.06 0.08 0.21 0.00 0.05 Th-228 Th-232 Note: Blank field indicates no reported result 32

BMRC Characterization Report Revision 0 4.5 Exterior Soils and Building Surfaces 4.5.1 Sub-Surface Borings A total of 19 sub-surface soil boring locations were identified in the immediate vicinity of the building exterior including the area surrounding the liquid waste tanks as shown in Figure 7.

One location (BMRC-018) was excluded because of its proximity to buried utilities.

Figure 7: Soil Bore Locations Borings were advanced by a split spoon sampler until refusal. Samples were collected in 2-foot intervals. All samples were field scanned at the time of collection for gross gamma radioactivity.

At a minimum the first (surface) and last (refusal) samples from each boring were sent for laboratory analysis. None of the collected samples indicated presence of groundwater.

The soil boring identified as BMRC-019 was advanced at the location of the former cooling tower which was known to have leaked low levels of radioactive liquids. The analyzed sample results only show the presence of naturally occurring radioactive materials.

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BMRC Characterization Report Revision 0 Examining the analytical data for the soil samples shows three (3) radionuclides as detectable.

These radionuclides are Cs-137, H-3, and Ni-63. Of these radionuclides, Ni-63 appears in the highest concentration with a maximum value of 11.9 pCi/g which is less than 1% of the NRC screening level and approximately 1.4% of the scaled NY State screening level. In terms of percent of screening level, H-3 appears in one sample at approximately 4% of the NRC screening level which is approximately 9% of the scaled NY State screening level. A summary of the analytical results is provided in Table 4-9 on the following page.

4.5.2 Surface Water and Groundwater Samples No surface water or groundwater samples were collected during characterization. As identified in the NRCs Safety Evaluation Report, the SUNY UB campus is covered with a dense glacial clay overburden that is generally 10 to 20 meters thick. Below the overburden is bedrock consisting of a combination of two limestone formations above five dolomite formations on top of a shale layer. The bedrock layers are approximately 1000 feet in total depth. Additionally, the NRC Safety Evaluation Report for the BMRC describes the water movement in the overburden as extremely slow with percolation only in the top foot of soil.

During the characterization effort, borings advanced to the upper surface of bedrock returned no groundwater. The sample analyses showed only low levels of radionuclide concentrations which may be attributable to the facility.

There are no surface water features in the immediate vicinity of the reactor. The nearest surface water feature is a small pond 0.6 miles south of the BMRC located in McCarthy Park.

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BMRC Characterization Report Revision 0 Table 4-9 Soil Borings Analytical Results Summary Bore ID> BMRC-001 BMRC-002 BMRC-003 BMRC-004 BMRC-005 BMRC-006 Max Max Max Max Max Max Max Max Max Max Max Max Activity MDC Activity MDC Activity MDC Activity MDC Activity MDC Activity MDC Radionuclide (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g)

Ac-228 3.23 0.20 1.47 0.20 2.07 0.24 1.65 0.19 2.08 0.17 Ag-108M 0.01 0.04 0.02 0.05 0.00 0.04 0.00 0.04 0.02 0.10 0.00 0.04 Ag-110M 0.00 0.07 -0.01 0.08 0.03 0.08 0.01 0.07 0.02 0.16 -0.01 0.06 Co-58 0.01 0.16 -0.03 0.21 -0.04 0.20 0.01 0.18 -0.02 0.38 0.04 0.17 Co-60 0.03 0.06 0.02 0.06 0.03 0.07 0.02 0.06 0.02 0.11 -0.01 0.04 Cs-137 0.09 0.06 0.08 0.07 0.07 0.08 0.03 0.06 -0.01 0.12 0.07 0.05 Eu-152 0.04 0.16 0.00 0.15 0.04 0.16 0.10 0.14 0.08 0.30 0.06 0.13 Eu-154 0.03 0.10 0.03 0.10 -0.02 0.10 0.01 0.08 0.05 0.14 0.00 0.07 H-3 0.09 0.89 2.91 0.89 -0.31 0.88 3.72 0.88 -0.39 0.86 -0.69 0.89 K-40 19.46 0.39 17.55 0.46 15.92 0.47 17.77 0.40 19.73 1.25 15.46 0.40 La-140 10.06 67.08 10.81 63.54 40.84 90.58 27.19 87.05 10.64 202.10 8.39 74.83 Mn-54 0.02 0.07 0.03 0.08 -0.01 0.08 0.01 0.07 0.03 0.15 0.03 0.07 Ni-63 9.17 0.83 11.90 0.92 0.25 0.60 0.34 0.66 -0.15 0.87 0.05 0.89 Ra-226 2.24 1.08 1.99 1.18 1.22 0.73 1.82 1.02 1.63 1.17 1.94 0.89 Sb-124 0.02 0.23 0.05 0.25 -0.01 0.25 0.01 0.22 -0.02 0.51 0.10 0.20 Th-228 0.99 0.10 0.79 0.11 0.85 0.10 0.98 0.08 0.65 0.17 0.81 0.08 Th-232 0.73 0.19 0.72 0.22 0.62 0.23 0.73 0.17 0.67 0.23 0.66 0.16 GR-A 7.27 4.89 8.46 4.30 5.88 4.29 4.83 3.87 11.50 4.95 9.43 2.85 GR-B 40.00 3.00 36.80 2.92 34.70 2.91 28.60 2.64 35.60 3.00 38.00 2.86 Note: Blank field indicates no reported result 35

BMRC Characterization Report Revision 0 Table 4-9 Soil Borings Analytical Results Summary (Cont.)

Bore ID> BMRC-007 BMRC-007A BMRC-008 BMRC-009 BMRC-010 BMRC-011 BMRC-012 Max Max Max Max Max Max Max Max Max Max Max Max Max Max Activity MDC Activity MDC Activity MDC Activity MDC Activity MDC Activity MDC Activity MDC Radionuclide (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g)

Ac-228 1.00 0.12 1.98 0.17 2.73 0.52 1.54 0.14 1.19 0.18 Ag-108M 0.01 0.03 0.00 0.05 0.03 0.12 0.01 0.04 0.01 0.05 0.00 0.05 0.00 0.05 Ag-110M 0.02 0.05 0.03 0.08 0.02 0.20 0.03 0.07 0.02 0.08 0.03 0.07 0.00 0.07 Co-58 0.02 0.12 0.02 0.24 0.02 0.47 0.02 0.17 0.02 0.18 -0.04 0.18 -0.03 0.20 Co-60 0.00 0.04 0.03 0.07 0.02 0.11 0.02 0.05 0.04 0.06 0.03 0.06 0.00 0.06 Cs-137 0.00 0.04 -0.02 0.07 0.09 0.16 0.05 0.05 0.08 0.06 0.05 0.06 0.04 0.06 Eu-152 0.03 0.09 0.00 0.18 0.10 0.39 0.06 0.13 0.05 0.17 0.01 0.14 0.04 0.16 Eu-154 0.01 0.06 0.00 0.09 0.03 0.21 0.00 0.07 0.04 0.10 -0.01 0.09 0.01 0.10 H-3 -0.46 0.78 4.12 0.88 -0.18 0.86 -0.24 0.81 -0.58 0.83 -0.55 0.84 -0.40 0.87 K-40 14.85 0.27 15.10 0.65 17.93 1.40 17.13 0.43 15.22 0.46 19.14 0.45 15.87 0.50 La-140 -23.60 53.96 5.54 115.00 28.14 283.90 13.33 85.31 19.13 79.09 -0.65 107.00 -11.06 99.19 Mn-54 0.02 0.05 0.01 0.09 0.04 0.19 0.01 0.07 0.00 0.08 0.03 0.08 0.03 0.08 Ni-63 1.67 0.93 0.46 0.65 0.18 0.66 0.25 0.68 0.36 0.66 0.51 0.70 0.60 0.71 Ra-226 1.70 0.93 1.41 0.84 1.70 1.10 1.95 1.12 Sb-124 0.10 0.16 0.04 0.30 -0.02 0.66 0.09 0.20 0.00 0.25 0.14 0.28 0.04 0.26 Th-228 0.59 0.06 0.75 0.10 0.92 0.28 0.81 0.08 0.76 0.10 0.88 0.10 0.80 0.11 Th-232 0.52 0.12 0.61 0.23 0.68 0.16 0.69 0.17 0.62 0.18 0.73 0.18 0.79 0.21 GR-A 4.92 2.81 7.56 2.71 5.48 2.72 7.60 3.06 5.55 2.89 7.72 2.91 9.77 3.27 GR-B 30.00 2.86 36.20 2.82 25.20 2.82 34.30 2.88 27.30 2.87 35.20 2.87 27.00 2.88 Note: Blank field indicates no reported result Note: BMRC-007 was moved due to concrete wall of N-16 tank room, the new hole was labeled BMRC-007A 36

BMRC Characterization Report Revision 0 Table 4-9 Soil Borings Analytical Results Summary (Cont.)

Bore ID> BMRC-013 BMRC-014 BMRC-015 BMRC-016 BMRC-017 BMRC-019 Max Max Max Max Max Max Max Max Max Max Max Max Activity MDC Activity MDC Activity MDC Activity MDC Activity MDC Activity MDC Radionuclide (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g)

AC-228 1.48 0.21 2.59 0.21 1.84 0.21 2.58 0.22 1.44 0.21 1.31 0.17 AG-108M 0.02 0.05 0.01 0.04 0.02 0.05 -0.01 0.04 -0.01 0.05 0.01 0.05 AG-110M 0.06 0.07 0.03 0.07 0.02 0.07 0.01 0.08 0.02 0.07 -0.01 0.06 CO-58 -0.03 0.18 -0.04 0.18 0.00 0.18 -0.01 0.18 0.01 0.19 0.01 0.15 CO-60 0.01 0.05 0.02 0.06 0.02 0.06 -0.01 0.05 0.02 0.07 0.01 0.05 CS-137 0.07 0.07 0.01 0.06 0.07 0.07 0.10 0.06 0.08 0.06 0.01 0.05 EU-152 0.02 0.15 -0.01 0.15 0.03 0.15 0.06 0.15 0.00 0.15 0.07 0.14 EU-154 0.02 0.09 0.02 0.09 -0.01 0.09 0.01 0.09 0.01 0.09 0.02 0.08 H-3 -0.48 0.83 0.38 0.97 0.17 0.99 0.19 0.97 0.29 0.95 0.16 1.07 K-40 16.27 0.46 15.37 0.41 17.76 0.40 15.34 0.40 18.29 0.50 18.38 0.38 LA-140 40.73 110.30 46.46 99.35 9.87 85.15 -8.27 76.33 12.88 79.46 15.03 75.23 MN-54 0.02 0.07 0.03 0.07 0.01 0.07 0.01 0.07 0.02 0.07 0.02 0.06 NI-63 0.31 0.66 -0.12 1.53 -0.39 1.53 0.00 1.55 -0.21 1.60 -0.24 1.57 RA-226 2.17 1.06 2.17 0.93 1.68 0.88 2.23 1.10 1.69 0.77 SB-124 0.02 0.24 0.01 0.24 0.03 0.24 -0.02 0.24 0.08 0.25 0.17 0.23 TH-228 0.79 0.09 0.75 0.13 0.85 0.10 0.91 0.09 0.86 0.10 0.73 0.09 TH-232 0.59 0.20 0.70 0.19 0.86 0.19 0.66 0.20 0.78 0.20 0.63 0.16 GR-A 5.98 3.12 6.21 2.86 8.14 2.88 7.59 3.45 7.29 3.23 8.62 3.01 GR-B 31.10 2.88 29.90 2.87 37.50 2.87 35.10 2.88 29.90 2.88 28.90 2.88 Note: Blank field indicates no reported result Note: BMRC-018 was excluded due to proximity to buried utilities 37

BMRC Characterization Report Revision 0 4.5.3 Gamma Radiation Scan A Gamma Radiation Scan was performed on the surface soils immediately surrounding the BMRC using a 2x2 Sodium Iodide (NaI) detector connected to a Ludlum Model 2221. The ratemeter reading was paired with its latitude/longitude coordinate and logged using a Trimble GPS unit. Additionally, an area several hundred yards to the northeast was selected for background radiation scans.

The average background reading was 6384 cpm with a maximum reading of 7311 cpm. The average reading for the surface soils near the BMRC was 6998 cpm with a maximum reading of 8909 cpm. The highest measurements recorded are adjacent to the current Vertical 10K Storage Tank and next to the northeast corner of the building. The elevated readings can be attributed to background interference from the Vertical 10K tank and the Cs-137 calibration source in Room 115A, respectively.

Figure 8: Gamma Walkover Data Value

>12768 12500 12000 11500 11000 10500 10000 9500 9000 8500 8000 7500 STEAM M.H.

7000 6500 6000 5500 5000 38

BMRC Characterization Report Revision 0 4.5.4 BMRC Rooftops The rooftop of the BMRC Containment Building and the Administrative Wing were surveyed for loose alpha/beta, fixed beta and loose beta. Additionally, the fixtures of the building exhaust systems were surveyed. Fourteen (14) random locations were plotted using the VSP software and an additional 6 locations were chosen based on professional judgment which included the Containment Exhaust Vent, the Tank Farm Vent pipe, and four vents from the Administrative Wing. All smear results were less than MDC. The static Beta measurements, without material specific background subtraction, had an average of 1369 dpm/100cm2, with a maximum result of 2403 dpm/100cm2 and a standard deviation of 456 dpm/100cm2.

4.5.5 Tritium and Carbon-14 Smears A tritium and Carbon-14 smear was collected from every room, sink, and hood throughout the BMRC Administrative Wing and Containment Buildings. A total of 83 smears (60 floor locations, 11 hoods, and 12 sinks) were analyzed by the offsite laboratory for loose tritium and loose C-14 with an MDC values at approximately 9 dpm/100 cm2 and 6 dpm/100 cm2, respectively. All locations in the Administrative Wing were less than MDC for both radionuclides. Eight (8) locations in the Containment Building indicated a potential presence of tritium with an average result of 30 dpm/100 cm2 and a maximum value of 100 dpm/100 cm2.

Additionally, three (3) locations indicated the potential presence of C-14 with a maximum result of 35 dpm/100 cm2.

When compared to the NRC Building Surfaces screening values for tritium and C-14, the maximum results are approximately 0.00005% and 0.0009% of the screening values, respectively.

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BMRC Characterization Report Revision 0 4.6 Tank Farm The Tank Farm lies in the subsurface adjacent to the containment building and extends towards the southwest of the containment building. The Tank Farm consists of two 250-gallon stainless steel tanks (Tanks 4 and 5) and two 600-gallon stainless steel tanks (Tanks 2 and 3) in connected concrete enclosures. There is also a 10,000-gallon carbon steel tank (Tank 1) in a separate enclosure to the southeast as shown in the HSA. The 250-gallon tanks were connected to the Isotope Processing Labs and the Hot Cell. The 600-gallon tanks supported the drains from the lower level labs and the floor drains. The tanks were interconnected by a manifold and pump system that allowed any combination of water transfer. The normal movement was to pump the small tanks to the 10,000-gallon tank for blending prior to discharges into the sewer system.

Samples from the all of the tanks were taken and sent for analysis. The results are contained in Table 4-10 and Figure 9 provides the orientation of the tanks in the tank farm.

As discussed in the HSA, the tanks were placed in wet layup by pumping them at least half-full of water to minimize oxygen degradation in 1985. This method of storage was recommended by personnel at the DOEs Hanford Site. Sampling ports known as Tell-Tale sampling ports were installed in the original configuration to sample any liquids that may have entered the enclosures and concrete berm to verify tank and piping integrity. BMRC personnel collect samples on a quarterly basis and no indications of loss of integrity have been detected.

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BMRC Characterization Report Revision 0 Figure 9: Tank Farm Layout 41

BMRC Characterization Report Revision 0 Table 4-10 Tank Farm Analytical Sample Results Summary BMRC-TF-001 BMRC-TF-002 BMRC-TF-003 BMRC-TF-004 BMRC-TF-005 Max Max Max Max Max Max Max Max Max Max Activity MDC Activity MDC Activity MDC Activity MDC Activity MDC Radionuclide (pCi/L) (pCi/L) (pCi/L) (pCi/L) (pCi/L) (pCi/L) (pCi/L) (pCi/L) (pCi/L) (pCi/L)

Ag-108M 116 3 2456 18 244 8 59 7 660 9 Ag-110M 2 3 377 27 -1 9 1 9 2 8 Am-241 (AS) 0 0 C-14 422 36 1510 25 1030 25 87 29 85 19 Cm-242 (AS) 0 0 Cm-243/244 0 0 (AS)

Co-58 0 3 -4 28 -1 12 -5 12 3 11 Co-60 128 3 10490 14 1654 8 2358 7 943 7 Cs-134 -1 3 1 19 -5 8 2 8 -7 8 Cs-137 80 3 437 22 221 10 130 9 105 9 Eu-152 1 9 -53 51 -10 25 0 21 -3 25 Eu-154 1 6 -8 24 -5 14 -4 10 6 13 Fe-55 0 216 GR-A 11 3 8 3 9 3 3 3 2 3 GR-B 10500 21 4780 15 5410 16 3180 12 2240 10 H-3 4320 3550 5810 3580 4260 3560 3880 3620 3970 3610 I-129 -30 77 La-140 0 6 -4 25 -13 14 -2 18 3 16 Mn-54 1 3 -2 26 3 10 2 11 -2 10 Ni-59 -47 73 Ni-63 388 14 10400 13 25500 13 1650 13 467 13 Pu-238 (AS) 0 0 Pu-239/240 0 0 (AS)

Pu-241 66 69 Sb-124 0 4 -8 22 -5 10 -5 10 2 10 Sr-89 13800 69 4220 64 5430 62 3720 63 2960 70 Sr-90 4310 18 1330 16 2300 15 1150 16 895 12 Tc-99 4 18 U-233/234 1 0 (AS)

U-235 (AS) 0 0 U-238 (AS) 0 0 Note: Blank field indicates no reported result. (AS) = Alpha Spectroscopy.

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BMRC Characterization Report Revision 0 4.7 Ventilation Surveys of the ventilation systems consisted of removable activity surveys at select access points and were performed in the course of the characterization surveys and are reported for the survey unit to which the system is associated. Table 4-11 provides a summary of the ventilation systems by survey unit.

Table 4-11 Ventilation Systems Survey Results Summary Removable Alpha Activity Removable Beta activity (DPM/100cm2) (DPM/100cm2)

Survey Unit Average Max Average Max 1C3 3 3 43 74 1G12 0 0 0 0 1G8 6 12 3 5 1G9 0 0 9 27 2C1 1 3 27 90 2G1 0 3 12 74 4.8 Miscellaneous Sample Areas Additional samples have been collected from the N16 tank room soil and from the sediment material in the 1K tank located in Room N03.

4.8.1 N16 Tank Vault The N16 tank is located in a vault in the sub-basement accessible through a wall hatch in Room N03. This room is unfinished and has concrete walls and a dirt floor. The tank is mounted on concrete piers. Piping exits the neutron level of containment and passes into this room.

Four (4) soil samples were collected as part of the characterization as shown in Figure 10. The results indicate impacts greater than the NRC and NYS Screening values for Co-60.

Additionally, Ag-108m was elevated and has a high potential to be greater than the release criteria developed as part of the DP. A summary of the analytical sample results is presented in Table 4-12 on the following page.

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BMRC Characterization Report Revision 0 Figure 10: N16 Tank Vault Sample Locations Table 4-12 N16 Tank Vault Analytical Results Summary Maximum Maximum Activity Maximum Activity Maximum Radionuclide (pCi/g) MDC (pCi/g) Radionuclide (pCi/g) MDC (pCi/g)

Ac-228 3.95E+00 4.71E-01 Fe-55 2.61E+00 9.81E+00 Ag-108M 9.99E+01 1.94E-01 H-3 7.19E+00 1.18E+00 Ag-110M 5.98E-01 1.42E-01 K-40 1.12E+01 4.60E-01 Co-57 9.13E-01 1.04E-01 La-140 1.84E-01 2.52E-01 Co-58 2.84E-02 1.60E-01 Mn-54 2.31E-02 1.29E-01 Co-60 1.26E+01 7.03E-02 Ni-63 7.69E+01 6.26E-01 Cs-134 3.03E-02 1.46E-01 Sb-124 2.50E-02 1.95E-01 Cs-137 4.65E+00 1.42E-01 Sr-89 5.66E-02 1.79E-01 Eu-152 2.00E+00 5.65E-01 Sr-90 3.61E-01 6.76E-02 Eu-154 9.50E-01 2.94E-01 Th-228 5.38E-01 3.28E-01 GR-A 8.36E+00 2.64E+00 Th-232 6.12E-01 4.82E-01 GR-B 8.27E+01 2.60E+00 44

BMRC Characterization Report Revision 0 4.8.2 1K Tank The 1K tank is located in a pit in Room N03. Water is collected in this tank from all sumps and drains in the containment building prior to entering the 10K tank and ultimately being discharged through the NPDES permitted sanitary system. The tank sediments were sampled and contained elevated levels of the site specific ROCs with the primary radionuclides being Co-60, Ag-108m, Ni-63, and Sr-90. The sediment sample collected from the tank had a dose rate of 1.5 mR/hr on contact. The water in the tanks was sampled and contained only Ni-63.

Table 4-13 on the following page provides a summary of the analytical results of the tank sediment.

Table 4-13 1K Tank Analytical Results Summary Sediments Liquid Uncertainty Uncertainty Activity Activity 2 Sigma 2 Sigma (pCi/g) (pCi/l)

Radionuclide (pCi/g) (pCi/l)

C-14 2.23E+01 6.70E+00 <MDC <MDC Ni-59 1.76E+02 1.03E+02 <MDC <MDC Ni-63 9.92E+03 9.94E+01 2.96E+03 4.29E+02 Sr-89 1.38E+02 2.03E+01 <MDC <MDC Sr-90 1.99E+02 1.37E+01 <MDC <MDC Ag-108m 1.03E+04 5.97E+01 <MDC <MDC Co-60 1.51E+03 3.57E+01 <MDC <MDC Cs-137 4.90E+01 3.12E+01 <MDC <MDC Am-241 3.90E+00 9.07E-01 <MDC <MDC Pu-238 1.71E+00 4.28E-01 <MDC <MDC Pu-239/240 5.74E+00 8.30E-01 <MDC <MDC U-233/234 1.74E+00 4.80E-01 <MDC <MDC U-238 1.09E+00 3.51E-01 <MDC <MDC 45

BMRC Characterization Report Revision 0

5. HAZARDOUS MATERIALS As part of the facility characterization, a hazardous materials assessment was also performed.

The goal of the assessment was to identify known and potential hazardous materials present at the facility. The assessment was based on site observations and document reviews.

During the on-site characterization, several potential hazardous materials were identified. These included caustics and acids (water treatment chemicals) and lubricating oils. There appeared to be no evidence of spill or leaks of these materials that would have impacted the facility. Because the facility housed many experiments, there may also be containers of laboratory chemicals stored about the facility that were not observed during the radiological site characterization activities. It is expected that containers of hazardous materials will be disposed of according to current practices and procedures prior to radiological decontamination and demolition.

UB does not have a site asbestos management plan, however, the facility has been surveyed for Asbestos Containing Materials (ACM) with ACM locations identified on piping. The floor tile in the administration wing is also known to be ACM. There is a high potential that the mastic under the tile is also ACM. The University plans on issuing a request for proposal (RFP) for asbestos removal in the second half of 2011. A new ACM survey will be performed before an RFP is issued for the ACM interference removal project.

Lead is present throughout the facility in the form of lead bricks and lead sheets. The University is in the process of issuing a contract to dispose of all loose legacy materials in the facility which includes all lead bricks and sheets. While a lead paint survey was not conducted, because of the age of the facility, it is expected that lead-based paints potentially were used throughout the facility. A lead paint survey will either be conducted at the same time a new ACM survey is conducted or will be conducted prior to an RFP being developed for building remediation and demolition.

Mercury chloride was used in the activation analysis lab on the gamma deck in the containment wing. The University required mercury monitoring during radiological surveys in the lab.

Additionally, the fluorescent light ballasts and any old thermostats will likely contain hazardous materials.

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6. QUALITY ASSURANCE AND QUALITY CONTROL ENERCON ensured that quality and regulatory requirements were satisfied during the characterization effort. Activities were controlled by procedures and the Site Characterization Plan. These documents included the following Quality Control (QC) measures as an integral part of the survey process.

6.1 General Provisions 6.1.1 Written Procedures Survey tasks were controlled by characterization survey instructions contained in each survey package, ENERCON procedures, and the BMRC Characterization Plan. The following procedures were used and referenced in the characterization survey instruction during the characterization project:

x TS 5.4.6, Chain of Custody x TS 5.4.7, Documentation of Radiological Surveys x TS 5.4.8, Instrument Source and Response Check x TS 5.4.10, Sampling Materials for Radiological Characterization x TS 5.4.11, Sample Handling x TS 5.4.12, Radiological Scans and Measurements x TS 5.4.13, Technician Training and Qualification x SUNY-OP-01, BMRC Reactor Bioshield Core Sampling x SUNY-OP-02, Neutron Deck Coring and Subgrade Sampling x SUNY-OP-03, Concrete Core Sample Processing x SUNY-OP-04, Soil Sampling 6.1.2 Instrumentation Selection, Calibration, and Use ENERCON selected instruments that are proven to reliably detect the radionuclides present at the BMRC. Instruments were calibrated by qualified vendors under approved procedures using calibration sources traceable to the National Institute of Standards and Technology (NIST). All detectors were subject to daily response checks when in use per guidance established in American National Standard Institute (ANSI) standard ANSI N323-1997.

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BMRC Characterization Report Revision 0 6.1.3 Chain of Custody The Chain of Custody procedure established responsibility for the custody of samples from the time of collection until results were obtained. All samples shipped off site for analysis were accompanied by a chain-of-custody record to track each sample.

6.1.4 Independent Review of Survey Results The survey package and survey data from each area received an independent review to verify all documentation is complete and accurate.

6.2 Training All project personnel received site specific training to identify the specific hazards present in the survey areas in addition to the characterization process. The training included a briefing and review of the BMRC Characterization Plan, ENERCON procedures, and the Site Safety and Health Plan. Personnel were also required to read each document. Copies of all training records were maintained on site through the duration of the onsite activities.

6.3 Sample Analysis ENERCON utilized a third party vendor, Teledyne Brown Engineering (TBE) in Knoxville, TN, for all radioanalytical services during the characterization project. All samples sent to TBE were accompanied by a Chain of Custody with their receipt acknowledge via email. TBE is used extensively in the nuclear industry for environmental, bioassay, and waste characterization analyses. TBE provides QC with all laboratory reports including blanks, spikes and duplicates.

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7. LIMITATIONS 7.1 High Background Areas As noted throughout this report, there were several areas where the background radiation was elevated due to the presence of stored radioactive materials. The status of these rooms at the time of the survey is documented in the photographs in Appendix A. In these areas, surveys may not be complete, but will be verified following the pre-decommissioning cleanout of loose legacy materials. Following the disposal campaign, radiological surveys will be performed in the areas and the results documented. The results will be analyzed to ensure the information submitted in the DP remains valid. A revised DP will be submitted to the NRC if a revision is required.

7.2 Internal Activity The internal surfaces of various components potentially containing radioactive materials were not surveyed. These include the primary coolant pipes, the nitrogen-16 decay tanks, the primary coolant pump, the vertical 10K taste tank, and the reactor tank. However, external dose rates were measured for these items and were documented in the radiation surveys. A dose rate survey was conducted in the reactor tanks near all activated components. This information was used by WMG for the Component Activation Analysis report.

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8. REFERENCES 8.1 BMRC Historical Site Assessment, Rev. 0, March 17 2010, ENERCON 8.2 BMRC Characterization Plan, November 24, 2010, ENERCON 8.3 BMRC Site Health and Safety Plan for the SUNY-UB Characterization, November 30, 2010, ENERCON 8.4 NUREG-1537, Part 1, Chapter 17, Guidelines for Preparing and Reviewing Applications for Licensing of Non-Power Reactors 8.5 NUREG-1757, Consolidated Decommissioning Guidance 8.6 NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM).

8.7 10CFR20, Standards for the protection against Radiation 8.8 DSHM-RAD-05-01, Cleanup Guidelines for Soils Contaminated with Radioactive Materials, New York State Department of Environmental Conservation.

8.9 ANSI N323-1997, Radiation Protection Instrumentation Test and Calibration, Portable Survey Instruments, American National Standard Institute (ANSI) Standard 8.10 NUREG-0982, Safety Evaluation Report related to the renewal of the operating license for the Research Reactor at the State University of New York at Buffalo, May 1983, NRC 8.11 NUREG/CR-3474, Long-Lived Activation Products in Reactor Materials.

8.12 NUREG/CR-4289, Residual Radionuclide Contamination Within and Around Commercial Nuclear Power Plants.

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BMRC Characterization Report Revision 0 Appendix A BMRC Characterization Photographs 51

BMRC Characterization Report Revision 0 Administrative Wing - Control Deck Level Photograph 1 Building Exterior Main Entrance Photograph 4 Room 213 Photograph 2 OTB (Vertical Accelerator Room) Photograph 5 Stairwell (S240)

Photograph 3 Room 215 Photograph 6 Room 221 (Women's Restroom) 52

BMRC Characterization Report Revision 0 Photograph 7 Main Control Deck Hallway Photograph 10 Room 214 (C241)

Photograph 8 Room 209 (Men's Restroom) Photograph 11 Room 210/212 Photograph 9 Room 216 Photograph 12 Room 208 53

BMRC Characterization Report Revision 0 Photograph 13 Room 208A 54

BMRC Characterization Report Revision 0 Administrative Wing - Gamma Deck Level Photograph 14 Room 115A (Vertical Photograph 17 Room 113 (Machine Shop)

Accelerator Target)

Photograph 18 Room 111 Photograph 15 Room 115 Photograph 19 Main Gamma Deck (C119)

Photograph 16 Stairwell (S122) 55

BMRC Characterization Report Revision 0 Photograph 20 Room 118 Photograph 22 Room 114 Photograph 21 Room 117 Photograph 23 Room 110 56

BMRC Characterization Report Revision 0 Administrative Wing - Neutron Deck Level (Sub-Basement)

Photograph 24 10K Tank Room Photograph 27 Room N02 Photograph 25 Room N05 Photograph 28 Room N04 Photograph 26 Room SN06 (Stairwell) Photograph 29 Room N03 57

BMRC Characterization Report Revision 0 Photograph 30 Room N03 1K Tank Pit Photograph 31 N16 Tank Room 58

BMRC Characterization Report Revision 0 Containment Wing - Control Deck Level Photograph 32 Room 201A Photograph 35 Room 204 (Fan Room)

Photograph 33 Room 202 Photograph 36 Room 205 (Control Room)

Photograph 34 Room 203 Photograph 37 Room 206 59

BMRC Characterization Report Revision 0 Photograph 38 Room 207 Photograph 41 Main Floor Control Deck Photograph 39 Room 200 Photograph 42 Main Floor Control Deck Photograph 40 Main Floor Control Deck Photograph 43 Main Floor Control Deck 60

BMRC Characterization Report Revision 0 Containment Wing - Gamma Deck Level Photograph 44 Room 101 Photograph 47 Room 109 Photograph 45 Room 103 Photograph 48 Room 108 Photograph 46 Room 104 Photograph 49 Room 107 61

BMRC Characterization Report Revision 0 Photograph 50 Room 106 Photograph 53 Main Floor Gamma Deck Photograph 51 Room 105 Photograph 54 Main Floor Gamma Deck Photograph 52 Hotcell Photograph 55 Main Floor Gamma Deck 62

BMRC Characterization Report Revision 0 Photograph 56 Main Floor Gamma Deck Photograph 57 Main Floor Gamma Deck 63

BMRC Characterization Report Revision 0 Containment Wing - Neutron Deck Level Photograph 58 Main Floor Neutron Level Photograph 61 Main Floor Neutron Level Photograph 59 Main Floor Neutron Level Photograph 62 Main Floor Neutron Level Photograph 60 Main Floor Neutron Level Photograph 63 Main Floor Neutron Level 64

BMRC Characterization Report Revision 0 Photograph 64 Main Floor Neutron Level Photograph 67 Thermal Column Photograph 65 Beam Port Storage Tubes Photograph 68 Trench System Photograph 66 High Radiation Area Neutron Level 65

BMRC Decommissioning Plan Revision 0 APPENDIX C - DECOMMISSIONING ENVIRONMENTAL REPORT

Decommissioning Environmental Report Buffalo Materials Research Center Prepared for:

Buffalo Material Research Center Office of Environment, Health, and Safety Services Completed by:

4490 Old William Penn Highway Murrysville, PA 15668 December 2011

BMRC Decommissioning Environmental Report Revision 0 Prepared By: Date: 11/4/2011 Wyatt Grant, P.E.

Senior Technical Specialist Prepared By: Date: 11/4/2011 Glenn Hargrove, PG Technical Specialist Reviewed By: Date: 11/4/2011 Robert D. Sanders Senior Technical Specialist Approved By: Date: 12/22/2011 Dustin G. Miller, CHP Decommissioning Project Manager

BMRC Decommissioning Environmental Report Revision 0 TABLE OF CONTENTS

1.0 INTRODUCTION

............................................................................................................................. 1 1.1 Purpose and Need for Action ......................................................................................................... 1 1.2 The Proposed Action ...................................................................................................................... 2 1.3 Applicable Regulatory Requirements, Permits, and Required Consultations ................................ 2 1.3.1 Federal Requirements.............................................................................................................. 2 1.3.2 State of New York................................................................................................................... 4 2.0 FACILITY DESCRIPTION............................................................................................................... 6 3.0 ALTERNATIVES............................................................................................................................ 13 3.1.1 SAFSTOR Alternative .......................................................................................................... 13 3.1.2 ENTOMB Alternative........................................................................................................... 13 3.1.3 DECON-A Alternative.......................................................................................................... 13 3.1.4 DECON-B Alternative .......................................................................................................... 14 3.2 Decommissioning of the Reactor, License Termination, and Site Reuse..................................... 14 3.2.1 Decommissioning of the Reactor .......................................................................................... 14 3.2.2 Licenses................................................................................................................................. 15 3.2.3 Reuse of Site ......................................................................................................................... 15 3.3 Cumulative Effects ....................................................................................................................... 16 3.4 Comparison of the Predicted Environmental Effects ................................................................... 16

4.0 DESCRIPTION

OF THE AFFECTED ENVIRONMENT ............................................................. 17 4.1 Land Use....................................................................................................................................... 17 4.2 Transportation .............................................................................................................................. 18 4.3 Geology and Soils......................................................................................................................... 18 4.4 Water Resources........................................................................................................................... 19 4.5 Ecological Resources.................................................................................................................... 20 4.6 Meteorology, Climatology, and Air Quality ................................................................................ 20 4.7 Noise............................................................................................................................................. 21 4.8 Cultural and Historical Resources ................................................................................................ 22 4.9 Visual/Scenic Resources .............................................................................................................. 23 4.10 Socioeconomic.......................................................................................................................... 23 4.11 Public and Occupational Health................................................................................................ 24 4.12 Waste Management................................................................................................................... 28 5.0 ENVIRONMENTAL IMPACTS..................................................................................................... 29 5.1 Land Use Impacts......................................................................................................................... 29 i

BMRC Decommissioning Environmental Report Revision 0 5.2 Transportation Impacts................................................................................................................. 29 5.3 Geology and Soils Impacts........................................................................................................... 30 5.4 Water Resources Impacts ............................................................................................................. 30 5.5 Ecological Resource Impacts ....................................................................................................... 30 5.6 Air Quality Impacts ...................................................................................................................... 30 5.7 Noise Impacts ............................................................................................................................... 31 5.8 Historical and Cultural Resources Impacts .................................................................................. 31 5.9 Visual/Scenic Resource Impacts .................................................................................................. 31 5.10 Socioeconomic Impacts ............................................................................................................ 31 5.11 Environmental Justice............................................................................................................... 32 5.12 Public and Occupational Health Impacts.................................................................................. 32 5.12.1 Nonradiological Impacts ....................................................................................................... 32 5.12.2 Radiological Impacts............................................................................................................. 32 5.13 Waste Management Impacts..................................................................................................... 33 5.13.1 Radiological Impacts............................................................................................................. 35 6.0 MITIGATION MEASURES ........................................................................................................... 36 7.0 ENVIRONMENTAL MONITORING AND MEASUREMENT PROGRAMS ............................ 37 7.1 Radiological Monitoring .............................................................................................................. 37 7.2 Physiochemical Monitoring.......................................................................................................... 37 7.3 Ecological Monitoring.................................................................................................................. 37 8.0 COST BENEFIT ANALYSIS ......................................................................................................... 38 9.0

SUMMARY

OF ENVIRONMENTAL CONSEQUENCES ........................................................... 39

10.0 REFERENCES

............................................................................................................................. 40 LIST OF FIGURES Figure 1: The BMRC and Adjacent Buildings.............................................................................................. 6 Figure 2: Control Deck Layout ..................................................................................................................... 9 Figure 3: Gamma Deck Layout................................................................................................................... 10 Figure 4: Neutron Deck Layout .................................................................................................................. 11 Figure 5: Decommissioning Boundary ....................................................................................................... 34 ii

BMRC Decommissioning Environmental Report Revision 0 Acronyms and Abbreviations ACM Asbestos Containing Material AST aboveground storage tank BLM Bureau of Land Management BMRC Buffalo Materials Research Center BSFR Bulk Survey for Release CAM Continuous Air Monitor CFR Code of Federal Regulations D&D Decontamination and Decommissioning DCGL Derived Concentration Guideline Levels DEC Department of Environmental Conservation (State of New York)

DECON-A complete decontamination and structure demolition option DECON-B complete decontamination and release of the structure option DOC Design and Oversight Contractor DOE United States Department of Energy DOT Department of Transportation ECL Environmental Conservation Law EH&S Environment, Health and Safety ENTOMB Entombment option EPA Environmental Protection Agency ER Environmental Report F Fahrenheit FSS Final Status Survey HEPA High-efficiency particulate air HSA Historical Site Assessment LLRW Low-level radioactive waste MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MGD Million Gallons per Day mrem Millirem MTR Materials Testing Reactor MW Megawatt MWt Megawatt thermal N-16 Nitrogen 16 NEPA National Environmental Policy Act iii

BMRC Decommissioning Environmental Report Revision 0 NMSS NRC Office of Nuclear Material Safety and Safeguards NRC Nuclear Regulatory Commission NSTC Nuclear Science and Technology Center NYCRR New York State Codes, Rules, and Regulations NYS New York State NYSDOH New York State Department of Health PuBe Plutonium-Beryllium PULSTAR Pulse Training Assembled Reactor RCRA Resource Conservation and Recovery Act RFT resilient floor tile SAFSTOR No-Action Alternative SCR Site Characterization Report (ENERCON initial draft)

SNM Special Nuclear Material Sr-90 Strontium-90 SUNY State University of New York TLD Thermoluminescent dosimeter UB University at Buffalo UST underground storage tank WMP Waste Management Plan iv

BMRC Decommissioning Environmental Report Revision 0

1.0 INTRODUCTION

This Environmental Report (ER) describes the environmental effects related to the decommissioning of the University at Buffalo (UB) Material Research Center (BMRC) Research and Test Reactor Facility.

The ER was prepared in accordance with the guidance provided in Chapter 6.0 of the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Material Safety and Safeguards (NMSS) NUREG-1748, Environmental Review Guidance for Licensing Actions Associated with NMSS Programs (NRC 2003b). This ER is designed to be used by the NRC in conducting its environmental assessment in accordance with the National Environmental Policy Act (NEPA) of 1969. NEPA requires federal agencies, as part of their decision-making process, to consider the environmental impacts of actions under their jurisdiction. The NRCs NEPA requirements are provided in Title 10 Code of Federal Regulations (CFR) Part 51.

1.1 Purpose and Need for Action The reactor at the BMRC was operated from 1961 to 1963 in support of training and education for the university and other national users. In 1964, the reactor was shut down and the core and control systems were modified so that the reactor could operate with Pulse Training Assembled Reactor (PULSTAR)-type fuel at power levels up to 2 megawatts thermal (MWt). The BMRC continued in support of training and education for the university from 1964 until reactor activities were suspended in 1994. During the period of operation, the BMRC has had multiple uses including:

x Training and education x Transient fuel performance testing x Materials radiation damage research x Isotope production x Neutron interrogation through activation analysis, radiography and delayed fission assay The unit has been in Possession Only status since June 6, 1997. All of the unused fuel was shipped to North Carolina State University in 1998 and all the spent fuel was removed from the site in 2005 and shipped to Idaho National Engineering and Environmental Laboratory. The BMRC administrative/laboratory wing (administrative wing) is currently used infrequently for two purposes. One purpose is an ion chamber calibration facility that is located in the basement of the administrative wing.

The second purpose is for training of the State University of New York (SUNY) at Buffalo (UB) maintenance staff in the proper care and cleaning of asbestos-containing resilient floor tile (RFT). The top 1

BMRC Decommissioning Environmental Report Revision 0 floor of the administrative wing has been determined to contain asbestos-containing RFT. Certain regulatory statutes mandate a standard of care for maintenance and custodial workers when activities affect asbestos-containing RFT. The UB Maintenance Department utilizes this space to train personnel on how to properly strip and wax the RFT. The containment building is only accessed for routine Technical Specification inspections.

1.2 The Proposed Action UB has already removed the BMRC from service and now plans to dismantle the reactor and its ancillary support systems, remove all residual radioactive material from the BMRC facility, and demolish the entire facility. Radioactivity levels will be reduced to levels that will permit release of the licensed area for unrestricted use and allow termination of NRC Possession License Only R-77.

1.3 Applicable Regulatory Requirements, Permits, and Required Consultations Decommissioning of the BMRC requires adherence to numerous federal, state, regional, local, and UB regulations. Guidance for determining many of the applicable federal, state, regional, and local requirements is identified in the following subsections. The information provided below is intended as a broad overview of applicable regulations and is not intended to be all-inclusive. The licensee or owner, UB in this case, is ultimately responsible for compliance with applicable federal, state, local, and internal regulations and is responsible for the costs associated with acquiring applicable permits and for the costs of implementing the necessary compliance programs during decommissioning activities.

1.3.1 Federal Requirements Decommissioning activities that are subject to federal regulations, permits, licenses, notifications, approvals, or acknowledgements include:

x Handling, packaging, and shipment of radioactive waste x Worker radiation protection x License termination and final site release x Worker, contractor, and the general publics health and safety x Liquid effluent releases x Hazardous waste generation and disposition x Handling, removal, and proper disposal of asbestos-containing materials x Handling, removal, and proper disposal of lead-containing paint and lead-based paint x Handling, removal and proper disposal of underground and aboveground storage tanks (UST and AST, respectively) 2

BMRC Decommissioning Environmental Report Revision 0 Nuclear Regulatory Commission The majority of radiological activities fall under Title 10 CFR and are administered by the NRC.

Applicable portions of Title 10 regulations are included within the following parts:

x Part 20 - Standards for Protection Against Radiation x Part 50 - Domestic Licensing of Production and Utilization Facilities including decommissioning activities x Part 51 - Environmental Protection Regulations For Domestic Licensing and Related Regulatory Functions x Part 61 - Licensing Requirements for Land Disposal of Radioactive Waste x Part 71 - Packaging and Transportation of Radioactive Material Many of the decommissioning requirements that involve activities for site control, characterization, and final status surveys (FSS) are found within the following parts of Title 10 of the CFR and are administered by the NRC. The parts include:

x Part 20.1401 - General provisions and scope x Part 20.1402 - Radiological criteria for unrestricted use x Part 20.1403 - Criteria for license termination under restricted conditions x Part 20.1404 - Alternate criteria for license termination x Part 20.1405 - Public notification and public participation x Part 20.1406 - Minimization of contamination x Part 20 Subpart F - Surveys and Monitoring x Part 30.36 - Expiration and termination of licenses and decommissioning of sites and separate buildings or outdoor areas x Part 40.42 - Expiration and termination of licenses and decommissioning of sites and separate buildings or outdoor areas x Part 70.38 - Expiration and termination of licenses and decommissioning of sites and separate buildings or outdoor areas x Part 72.54 - Expiration and termination of licenses and decommissioning of sites and separate buildings or outdoor areas Department of Transportation Radioactive material transportation activities fall under Title 49 of the CFR and are administered by the Department of Transportation (DOT). Applicable portions of Title 49 regulations are included within the following parts:

x Subtitle B - Other Regulations Relating to Transportation, Parts 100 to 185 - as applicable 3

BMRC Decommissioning Environmental Report Revision 0 Environmental Protection Agency The Environmental Protection Agency (EPA) provides the federal environmental requirements. The State of New York is under the jurisdiction of the EPA Region II office. The EPA regulations outlined in Title 40 of the CFR apply as follows:

x Part 61 - National Emission Standards for Hazardous Air Pollutants x Part 61 Subpart M - National Emission Standard for Asbestos pertaining to asbestos handling, removal, and disposal x Part 129 to 132 - Clean Water Act x Part 190 - Environmental Radiation Protection Standards for Nuclear Power Operations x Parts 260-272- Hazardous waste disposal and solid waste disposal as included in the Resource Conservation and Recovery Act (RCRA) 1.3.2 State of New York Environmental regulations and policies in the State of New York are codified by the New York State Legislature under Environmental Conservation Law (ECL) and enforced by the Department of Environmental Conservation (DEC). Portions of the ECL may apply as follows:

x Article 1- General Provisions o Title 1-Declaration of Policy x Article 3 - Department of Environmental Conservation; General Functions, Powers, Duties, and Jurisdiction x Article 15 - Water Resources o Title 1 - Short Title; Statement of Policy; Definitions; General Provisions o Title 5 - Protection of Water o Title 19 - Drainage o Title 31 - Groundwater Protection and Remediation Program x Article 17 - Water Pollution Control o Title 1 - General Provisions and Public Policy o Title 5 - Prohibitions o Title 7 - Permits and Certificates o Title 8 - State Pollution Discharge Elimination System o Title 14 - Nonpoint Source Water Pollution Control x Article 19 - Air Pollution Control o Short Title; Declarations of Policy and Purpose; Definitions x Article 27 - Collection, Treatment and Disposal of Refuse and Other Solid Waste 4

BMRC Decommissioning Environmental Report Revision 0 o Title 1 - Solid and Hazardous Waste Management Policy and Planning o Title 3 - Waste Transporter Permits Part 364 Waste Transporter Permits o Title 7 - Solid Waste Management and Resource Recovery Facilities o Title 9 - Industrial Hazardous Waste Management o Title 17 - Lead-Acid Battery Recycling o Title 24 - Environmental Tests Reporting Requirements x Article 28 - Pollution Prevention x Article 37 - Substances Hazardous or Acutely Hazardous to Public Health, Safety, or the Environment o Title 1 - Substances Hazardous to the Environment o Title 2 - Hazardous Packing x Article 54 - Environmental Protection Act o Title 1 - General Provisions x Article 72 - Environmental Regulatory Program Fees o Title 1 - Declaration of Policy; Definitions o Title 4 - Hazardous Waste Program Fee Worker protection laws in the State of New York are regulated and enforced by the State of New York Department of Labor. Portions of 12 New York State Codes, Rules, and Regulations (NYCRR) may apply as follows:

x Asbestos - Part 56 of Title 12 of the Official Compilation of Codes, Rules, and Regulations of the State of New York (cited as 12 NYCRR Part 56) 5

BMRC Decommissioning Environmental Report Revision 0 2.0 FACILITY DESCRIPTION The BMRC is owned by UB and is located on the south edge of the South Campus of UB in the City of Buffalo, New York. See Figure 2.1 for the location of the BMRC as it is positioned on the UB campus.

The facility consists of a cylindrical vapor containment building and the attached rectangular administrative wing that is oriented on the long axis to the northeast of the containment building. The reactor is housed within the containment building. The containment building is only accessed for routine technical specification inspections. An ion chamber calibration facility is located in the basement of the administrative wing and the top floor of the administrative wing is utilized as an area for training of UB maintenance and custodial staff.

Figure 1: The BMRC and Adjacent Buildings 6

BMRC Decommissioning Environmental Report Revision 0 The Containment Building The Reactor The containment building is a poured reinforced concrete right cylinder, 70 feet in diameter and 52 feet in height. The reinforced concrete walls and roof (supported by concrete beams) are 2 feet thick and 0.33 feet thick, respectively. The building walls and the first level are laid on bedrock. The containment building is attached to a laboratory complex formerly dedicated primarily to nuclear science-related research and instruction. The reactor pool is a reinforced high density concrete structure located mostly below grade.

The reactor facility, including the control and safety systems, was designed and built by American Machine and Foundry. The reactor first began operation in June 1961 and was originally referred to as the Western New York Nuclear Research Center and subsequently the Nuclear Science and Technology Center (NSTC) of SUNY UB. The reactor was originally fueled with materials-testing-reactor (MTR)-

type fuel elements operating at a maximum steady state power level of 1 MWt. The reactor operated in this mode from 1961 to 1964. The reactor was shut down in 1964 so that core and control systems could be modified. After the modification, the reactor operated with PULSTAR-type fuel at power levels up to 2 MWt. The original core grid plate was retained and the MTR fuel elements were replaced with PULSTAR pin-type fuel clusters which were designed to utilize the existing grid space. On May 12, 1965, the reactor was additionally licensed to operate in the pulse mode with a routine energy per pulse of up to 35 MW per second and a maximum size pulse of 44 MW per second. The BMRC operated with PULSTAR fuel since 1965, with one major re-fueling in 1978.

The reactor is a PULSTAR heterogeneous open-pool type water-cooled reactor using solid 6 percent enriched uranium fuel similar to nuclear power reactor fuel. The core was cooled by forced convective cooling at high power levels and by natural convection at lower power levels. The coolant/moderator is light water, and the reflector may be either water or graphite. The core is immersed in a 13,000-gallon, aluminum-lined reinforced concrete pool. The coolant was circulated through external systems for heat removal and for purification. Former reactor experimental facilities included in-core irradiation positions, a thermal column, beam tubes, pneumatic sample transport systems, a dry gamma chamber, and a gamma irradiation facility.

The core was immersed in a 13,000-gallon, aluminum lined tank pool surrounded by both high density and normal density concrete. It was cooled by a 5,000-gallon circulating water system exiting from the reactor pool bottom via pumps to a subsurface 5,000-gallon hold-up tank for N-16 decay. The water then went through a heat exchanger connected to an external cooling tower located on the southeast side of the 7

BMRC Decommissioning Environmental Report Revision 0 administrative wing. The cooled water was then circulated back to the top of the pool. After a leak was detected in 1977 where the coolant piping penetrated the bottom of the reactor pool, modifications were made to the cavity lower liner and the piping was rerouted to exit out of the side of the pool and through the lower wall of the Neutron Deck of the containment building. After a leak was detected in 1989 in the liner, a new liner was installed over the original liner in the lower section of the tank. All the spent and unused fuel has been shipped off-site. The reactor and all its associated components remain in the pool in the normal operational configuration as they were during operation.

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BMRC Decommissioning Environmental Report Revision 0 The Control Deck The Control Deck (top level) houses a dual-hook overhead crane with a ten-ton and a two-ton hook on tracks at the ceiling level. Crane access to the lower levels was via floor plates on the Control Deck and a webbing cover on a floor penetration from the Gamma Deck to the Neutron Deck. In addition to the reactor pool, the Control Deck houses the Reactor Control Room (Room 205), the Superintendents Office (Room 207), two labs (Rooms 202 and 203), the Fan Room (Room 204), another office (Room 206), and an airlock (Room 201B) for access to the upper or first floor of the administrative wing. The New Fuel Storage Room (Room 201A) is located between the inner and outer door of the airlock. Figure 2 below provides the layout of the Control Deck.

Figure 2: Control Deck Layout 9

BMRC Decommissioning Environmental Report Revision 0 The Gamma Deck The middle level of the Containment Building is called the Gamma Deck. A plan view of the Gamma Deck is shown on Figure 1-4. The Hot Cell, Hot Cell Work Room (Room 105), Locker Room (Room 106), Room 107, Hot Chemical Lab (Room 108), Medium Chemical Lab (Room 104), Activation Analysis Lab (Room 109), Lower Fuel Vault Room (Room 101A), and Chemical Lab (Room 103) are located on the Gamma Deck. This level also has an Airlock (Room 101) for access to the bottom level of the Administration Building and a truck door for access to the asphalt drive located on the southern side of the Containment Building.

The Truck Door provides vehicular and equipment access to the outside. Prior to 1979, the Truck Door area housed a waste compaction system. Wastes generated at the BMRC (NSTC at the time) and under the site-wide radiological program were collected and compacted for volume reduction in this system.

Figure 3 provides the layout of the Gamma Deck.

Figure 3: Gamma Deck Layout 10

BMRC Decommissioning Environmental Report Revision 0 The Neutron Deck The Neutron Deck (lower level) is divided by caging into two sections. The open area includes access to the Patient Treatment Room which is lined with wood. The Patient Treatment Room was included in the original design of the BMRC to be a location to provide neutron treatments to cancer patients. No history that it was ever used for this purpose was found. The enclosed area includes the rerouted Primary Coolant Piping and access to the seven Beam Storage Tubes. The Beam Storage Tubes extend 10 feet through the containment building wall into the surrounding soil. They were originally designed to hold the beam tube plugs but were used as storage for decay of high dose reactor components or experiments.

At the point where they exit the containment building wall, they are approximately 10 feet underground.

The balance of the Neutron Deck is open area where the majority of radioactive contaminated materials awaiting disposal are being stored. Figure 4 provides the Neutron Deck layout.

Figure 4: Neutron Deck Layout 11

BMRC Decommissioning Environmental Report Revision 0 The Administrative/Laboratory Wing The Upper Floor The administrative wing housed the Locker Rooms (Rooms 209, 221, and 221A), Conference Room (Room 212), offices (Rooms 208, 210, 214, and 216), the Reception Area (Room 208A), Upper Air Conditioning Equipment Room (Room 209B), the Janitors Closet (Room 209A) and two low level labs (Rooms 213 and 215) on the upper floor.

The Basement Floor On the lower or basement floor there was a Horizontal Accelerator (Room 118) with an adjacent cave area. This area has supported multiple functions; including a lab, a Class Room (Room 114), a Health Physicist Office (Room 117), Counting Room, The Vertical Accelerator Room (Rooms 115 and 115A),

and a Machine Shop (Room 113). The Electrical Service Room (Room 110) is also located on this level.

There is a tunnel off the Electrical Service Room (Room 110) that connects to the Campus Service Tunnel system and is locked to prevent access.

Subbasement The space below the basement floor level is the area designated as the subbasement. This area has rooms housing the N16 Vault (Room N16), the Cooling Water Equipment Room and the Facility Sumps (Room N03), the lower Air Conditioning Equipment Room (Room N02), the Air Handling Duct Room (Room N04), and access to the 10,000-gallon tank enclosure (Room N05). These rooms house the decay, filtration, and cooling systems for the reactor. Room N03 also contains the 1,000-gallon waste holding tank.

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BMRC Decommissioning Environmental Report Revision 0 3.0 ALTERNATIVES There are four alternatives available to, and considered by, the UB: 1) the No-Action alternative (SAFSTOR); 2) the entombment option (ENTOMB); 3) complete decontamination and structure demolition (DECON-A); and 4) complete decontamination and release of the structure (DECON-B). The DECON options are recommended by the NRC for non-power reactors. The selected alternative is DECON-A. The four alternatives are summarized in the following sections.

3.1.1 SAFSTOR Alternative The SAFSTOR alternative would involve the same potential risks and environmental impacts as the proposed action, but for a much greater time period. The alternative would require that the UB maintain current radiological controls, site security, required licensing, a reactor administrator, and the utilities until the DECON option was implemented. The UB will have to incur all of the expenses associated with maintaining the facility. This alternative would require that the UB apply for and obtain an extension to the current NRC operating license. The SAFSTOR alternative would leave the site as unusable space that will delay the beneficial reuse and the potential for environmental contamination would still exist. This alternative is not environmentally preferable.

3.1.2 ENTOMB Alternative The ENTOMB alternative would require the same as the SAFSTOR alternative with the additional requirement that the radioactive material is placed in a long-lived structure, i.e., concrete, or the current structure proven adequate for long-term storage. This alternative delays the impacts of the proposed action and the potential for environmental contamination would continue to exist. This alternative is not environmentally preferable.

3.1.3 DECON-A Alternative The facility records and current facility characterization reveal minimal facility contamination from past reactor operations. This alternative requires the site to be released and restored for unrestricted use. The reactor, the containment building, and the administrative building under this option will be disassembled and the radioactive material removed to meet the release criteria. The BMRC facility is located in the center of a growing university and the land area could be reused for future construction; therefore, complete decontamination and demolition (D&D) of the BMRC facility, DECON-A, is the preferred option. This alternative poses minimal risk and impacts to the environment as described in Section 8.0.

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BMRC Decommissioning Environmental Report Revision 0 3.1.4 DECON-B Alternative The DECON-B alternative is similar to DECON-A, however, the structure would not be removed after remediation. The risk and impacts to the environment are the same as DECON-A for this alternative, but the UB has determined that the reuse of the building is not in its best interests. Additionally, a final status survey of the soils and bed rock is more complicated and requires a significantly greater level of surveying and sampling than releasing an open excavation. This alternative is not preferred by the UB.

3.2 Decommissioning of the Reactor, License Termination, and Site Reuse 3.2.1 Decommissioning of the Reactor UB plans to decommission the reactor which will require the removal of all radioactive materials from the BMRC facility; dismantling the reactor, the reactor peripheral support systems, and support buildings; and removal of the UST and associated piping, materials associated with the previously demolished cooling tower, and any legacy waste.

Some of the reactor components and systems are either activated or contaminated and will need to be segregated from non-radiological components and surfaces. Components and systems that are designated as activated or impacted will be disposed of as low-level radioactive waste (LLRW). Building materials, such as the reactor tanks, will need to be evaluated for radiological activity and removed and disposed of according to their radiological status, as necessary.

The following are decommissioning tasks, which are necessary for site release. The sequence in which these tasks occur may vary:

x Pre-Decommissioning Cleanout - removal of loose legacy equipment and materials Isolation and removal of inactive systems x Removal of hazardous materials (i.e., lead) and abatement of asbestos containing material (ACM) x Perform supplementary characterization x Installation of temporary systems and preparation of the facility for decommissioning operations x Removal of the irradiated reactor components x Removal of and disposition of water in the reactor x Removal of and disposition of the reactor liner x Decontamination of concrete associated with reactor, bioshield, and hot and dry cell x Removal of the reactor, bioshield, and hot and dry cell x Segregation, packaging and shipment of materials according to radioactivity levels x Removal of auxiliary systems (rabbit system, water purification, ventilation) x Decontamination of building surfaces x Interior release survey using Reg Guide 1.86 limits x Demolition of the BMRC facility x Perform the FSS on BMRC footprint 14

BMRC Decommissioning Environmental Report Revision 0 x Submit required reports that demonstrates to the NRC that the facility meets the release requirements x Request license R-77 termination x Restore the site by backfilling the void created by removing below grade structures for future use by UB.

The BMRC has been in Possession Only status since June 1997 and the spent fuel was removed from the site in 2005. The on-site decommissioning tasks are expected to start date in the second half of 2012.

The FSS will be developed by the Design and Oversight Contractor (DOC) using the criteria provided in NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) (NRC 2000).

Radiological measurements identified in the Characterization Report indicated that the typical levels of residual radioactive materials on the building surfaces are less than release criteria established by the NRC in Regulatory Guide 1.86. Residual radioactive materials were only found in areas indicated by the Historical Site Assessment. Soil samples indicated that only limited remediation of soils will be required outside of the facility. Radionuclides above derived concentration guideline levels (DCGL) were detected in the N-16 Vault, which does not have a concrete floor covering the bedrock. The N16 Vault is identified on Figure 4. The structures and components that contain the majority of the residual radioactive materials are the activated reactor components and bioshield, the former and current liquid waste tanks, and the reactor water cooling system. Residual radioactive materials were also identified in the Hot Cell and laboratory fume hoods. Since all BMRC structures, systems, and components will be removed from the site, the FSS will only need to cover the exposed surface soils.

3.2.2 Licenses The BMRC maintains NRC license No. R-77 for Possession Only. Recently in 2011, the BMRCs Special Nuclear Material (SNM) License No. SNM-273, for possession of a Plutonium-Beryllium (PuBe) neutron source was terminated and the PuBe source transferred to the Universitys New York State Department of Health (NYSDOH) radioactive material license, No. 1051. The scope of the NYSDOH license is for possession of sources and for decommissioning.

The R-77 license is to be terminated as part of the decommissioning. The MARSSIM NUREG-1575 provides guidance to assemble a statistically accurate FSS plan to support the termination of the NRC license. The NYSDOH license will have all sources transferred to another NYSDOH license at UB. The SNM license will have a request for license termination submitted after the NYSDOH license is amended in order to transfer the PuBe source.

3.2.3 Reuse of Site The site is expected to be reused as either green space or a future road.

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BMRC Decommissioning Environmental Report Revision 0 3.3 Cumulative Effects The cumulative effects of the implementation of the proposed action will result in short-term cumulative impacts as discussed in Section 5.0 of this ER.

3.4 Comparison of the Predicted Environmental Effects The following table provides the predicted environmental effects and impacts of the alternatives that were evaluated for this ER. More detailed information on the expected impacts can be found in Section 5.0 of this ER.

Affected Environment No Action Proposed Action Land Use Possible Impact Possible Impact Transportation No Impacts Possible Impact Geology and Soils Possible Impact Possible Impact Water Resources No Impacts No Impacts Ecological Resources No Impacts No Impacts Meteorology, Climatology and Air Quality No Impacts Possible Impact Noise No Impacts Possible Impact Cultural and Historic Resources No Impacts No Impacts Visual/Scenic Resources No Impacts No Impacts Socioeconomic No Impacts No Impacts Environmental Justice No Impacts No Impacts Public and Occupational Health No Impacts Possible Impact Waste Management No Impacts Possible Impact 16

BMRC Decommissioning Environmental Report Revision 0

4.0 DESCRIPTION

OF THE AFFECTED ENVIRONMENT The following sections describe specific areas of the environment that may be affected because of the decommissioning activities.

4.1 Land Use The BMRC is a stand-alone structure located on the south edge of the South Campus of UB. The structure is comprised of two distinctively different architectural components, the containment building and the attached administrative wing.

The containment building is a three-story right cylinder constructed of concrete, 70 feet in diameter with a height of 52 feet. The reactor is located within the three-story cylindrical building. The former core was immersed in a 13,000-gallon, aluminum-lined tank pool surrounded by both high density and normal concrete. The cylindrical building walls and the lower level of the containment building rest directly on bedrock, the exterior walls are poured concrete finished with paint, and the roof of the cylindrical building is four-inch thick reinforced concrete supported by concrete beams. Interior floors are poured concrete with the Neutron Deck formed of two courses of concrete with a finish course of about 1.25 inch thickness, and the Gamma and Control Decks are monolithic. Select floors in the administrative building are formed of terrazzo, mosaic tile, or resilient floor tiles. Ceilings in select areas may be formed of acoustic tiles of varying sizes resting on a metal grid system suspended with metal hangers.

The administrative wing is comprised of three levels, the Upper Floor, the Basement Floor, and the Foundation Level. The administrative wing was originally constructed for office, laboratory, utility rooms, and classrooms for activities conducted with and/or for the reactor. The exterior walls are formed of load-bearing concrete masonry blocks and are finished with a brick façade. Interior floors primarily consist of nine-inch by nine-inch vinyl asbestos floor tiles adhered to the slab with adhesive. Walls are formed of painted masonry block and finished with paint.

The Tank Farm The Tank Farm lies in the subsurface area south of the containment building and extends towards the southeast of the containment building. The Tank Farm consists of two 250-gallon stainless steel tanks and two 600-gallon stainless steel tanks in connected concrete enclosures. There is also a 10,000-gallon carbon steel tank outside of the enclosures to the east. The 250-gallon tanks were connected to the Isotope Processing Labs and the Hot Cell and the 600-gallon tanks supported the drains from the lower level labs and the floor drains. The tanks were interconnected by a manifold and pump system that allowed any 17

BMRC Decommissioning Environmental Report Revision 0 combination of water transfer. The normal movement was to pump the small tanks to the 10,000-gallon tank prior to discharging to the sewer system. The 10,000-gallon tank sits on concrete saddles inside a concrete bermed area. The enclosure and the 10,000-gallon tank are beneath 3-4 feet of soil.

4.2 Transportation The BMRC is located in a relatively moderate to high traffic area in the northeastern section of the city limits of Buffalo, New York. Surrounding areas are largely residential. A recent traffic congestion study by INRIX, as reported by a local Buffalo television station, indicated that while traffic congestion in Buffalo is better than the national average, it has increased from 2009 to 2010 faster than the national average. In Section 3.0, Highway System Profile of a Niagara Frontier Urban Area Freight Transportation Truck study indicated that the region has a total of 3,675 miles of highway network, including major interstates, state routes, and local arterial roads. The major interstate highways that service Erie and Niagara counties included Interstate 90, Interstate 190, Interstate 290, and Interstate 990. The local streets will be utilized to transport radioactive waste and construction debris associated with the decommissioning and demolition of the BMRC. Interstates 90 and 190 are less than five miles from the site and will likely be utilized as transport routes to the eventual disposal destinations.

4.3 Geology and Soils The City of Buffalo is located within the Niagara Region, a section of a great plain that extends north to south from the Laurentian Highlands (Canadian Shield) approximately 161 kilometers north of Toronto, Ontario to the Southern Allegheny Plateau. This plain is a small portion of the Great Lakes low lands within which all of the Great Lakes are located. Within the Niagara Region is located the Niagara Escarpment, the weathered edge of an ancient sea bottom extending roughly from Watertown, New York westerly through Illinois and Wisconsin. The Escarpment is neither a fault line nor a rift line, but is instead an erosional feature classified as a cuesta, a ridge with a gentle slope on one side and a cliff on the other. (NUREG-0982)

A smaller escarpment, the Onondaga Escarpment, often referred to as the Onondaga Cuesta, trends east to west located along the northern shore of Lake Erie. The escarpment is capped with the Onondaga limestone. The SUNY campus is located on the extreme edge of the Onondaga Cuesta. The cuesta faces the northwest and has a local relief of 80 feet. The southeastern portion of the campus is located on the dip slope of the cuesta. In general, the area within which the City of Buffalo is located is very flat with the exception of the Onondaga Cuesta. As a result, local streams are slow moving and widely spaced.

(NUREG-0982) 18

BMRC Decommissioning Environmental Report Revision 0 The bedrock of the area consists of Silurian and Middle Devonian marine shale, dolomite and limestone with a southerly drop of 50 feet per mile. The rock formation is a joint system consisting of two joint sets arranged in a north-south east-west configuration intersecting at approximately 90 degrees. This is underlain by a similar marine series of Ordovician and Cambrian age to a depth of 3,000 feet that rests on a pre-Cambrian deposition. (NUREG-0982)

The overburden from the bedrock to the surface averages about 15 feet in the area of the BMRC. The soil proper occupies the first 14 inches with subsoil, which is made of dense boulder and clay extending uniformly to the bedrock. The soil from the surface to a depth of nine inches consists of moderately compacted, brown-gray, heavy, silty clay loam. Beneath this is a roughly five-inch stratum of compact dull yellowish-brown silty clay with a considerable amount of fine pebbles. The layer below this is made up of dull brown clastic, heavily compacted, silty clay containing crystalline pebbles dispersed throughout. This clay material extends to bedrock without any significant change. (NUREG-0982)

The data in this section was collected during initial licensing and design of the BMRC. The primary sources of information for this section is the reactors Safety Analysis Report and NUREG-0982, Safety Evaluation Report related to the renewal of the operating license for the Research Reactor at the State University of New York at Buffalo. More recent drilling on the south campus in 2008 and 2009, supports the above description of the area geology and soils.

4.4 Water Resources The City of Buffalo is located along the eastern end of Lake Erie. Water comes from Lake Erie into an intake out in the lake and flows through a 12-foot by 12-foot conduit to the Colonel Ward Water Treatment Plant and out to the city through 800 miles of pipe controlled by 25,000 valves. The system supplies an average of 99 million gallons per day to 82,000 service connections and 7,600 fire hydrants. A peak supply of 127.8 MGD was reached in 2001. This capacity is adequate to satisfy foreseeable future demands. Water quality is good and complies with all regulations.

Four miles east of the campus at Williamsville is Elliot Creek, which flows north to an unnamed creek that has its headwaters approximately two miles north of UB campus. There are no other streams in close proximity to the BMRC. The nearest water access is Lake Erie, which is five miles to the southwest of the BMRC site. The Niagara River, on which Niagara Falls lie, is 4.6 miles to the southwest while Lake Ontario is approximately 30 miles from the facility. These bodies of water are not anticipated to be impacted by any activities at the BMRC.

The UB campus is located along the edge of the Onondaga Cuesta, which trends from the southwest to the northeast. The southeastern portion of the campus is positioned on the down slope side of the cuesta.

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BMRC Decommissioning Environmental Report Revision 0 With the exception of the cuesta, the general area is relatively flat. As a result, streams are slow moving and widely spaced apart. In the immediate area of the BMRC, stormwater run-off flows generally towards the south where it is collected by a stormwater management system and eventually routed to the sewage treatment plant on campus. There are no streams in the immediate vicinity of the BMRC. The slope of the adjoining land and the drainage path near the building will not permit accumulation of water around the building. There are no credible paths for reactor pool water to get into the campus water system, however surface water run-off is routed towards the UB sanitary sewer system. There are no surface water features in the immediate vicinity of the reactor. The nearest surface water feature is a small pond 0.6 miles south of the BMRC located in McCarthy Park.

The density of the clay soils does not allow percolation of water into the sub-surface. For all practical purposes, surface water penetration ceases below a depth of 13 inches. Water will move through the saturated and undisturbed material in the first eight inches of topsoil at an average rate of 0.02 to 0.2 inches per hour. In the next five-inch layer, the penetration decreases sharply to zero. The underlying Onondaga and substrata are not known to contain water in the immediate vicinity, but it is not impervious. Based on observations several miles east of the campus, the water was presumed to move to the escarpment edge. Water movement within the overburden including the soil proper is extremely slow.

Only within the first 13 inches water percolation takes place very slowly, 0.2 inches per hour. In the next 15 feet, permeability ceases for all practical purposes. Surface water flows in a southerly direction into established storm drains. (NUREG-0982) 4.5 Ecological Resources The BMRC is located within a well-developed area where virtually all flora consist of species introduced for landscaping of residential, commercial, and educational facilities. The likelihood of endangered native species of plants is very small. Any native fauna would consist primarily of small, nocturnal animals not likely to appear on any endangered species list. The DEC website for endangered species did not list any insects as endangered or threatened in Buffalo, New York.

4.6 Meteorology, Climatology, and Air Quality Buffalo experiences a fairly humid, continental- type climate, but with a definite "maritime" flavor due to strong modification from the Great Lakes. Lake Erie lies to the southwest of the facility and functions to moderate the temperatures. The average temperatures measured at the Buffalo International Airport range from 18 degrees to 80 degrees Fahrenheit (F) with extremes recorded of -20 degrees and 99 degrees F.

Annual rainfall averages 38.5 inches in the Buffalo area and snowfall averages 93.3 inches per year.

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BMRC Decommissioning Environmental Report Revision 0 Winters in Western New York are generally cloudy, cold, and snowy but are changeable and include frequent thaws and rain as well. Over half of the annual snowfall comes from the "lake-effect" process and is localized. Due to the prevailing winds, areas south of Buffalo receive much more lake-effect snow than locations to the north.

As the prevailing airflow is mostly southwesterly, areas near the lake are often as much as 20 degrees colder than inland locations. The cool lake waters act as a strong stabilizing influence, so areas near the lakeshore, which includes the City of Buffalo, experience more sunshine and fewer thunderstorms than inland areas.

Summer is also influenced by the stabilizing effect of Lake Erie. It inhibits thunderstorms and enhances sunshine in the immediate Buffalo area. The lake also modifies the extreme heat that approaches from the Ohio Valley. Tornados are rare in Buffalo; the Weather Channel website reports that only 19 tornados have struck the vicinity of Buffalo from the period 1950 to 2009.

4.7 Noise The BMRC is located within an academic setting and noise is typically at a low level. At the time of the drafting of this document, renovation activities were occurring at a nearby structure. The Environment, Health and Safety Department (EH&S) of UB SUNY has issued guidance for noise in their Contractor Safety Guidebook, Document Number SA-023-A, August 25, 2009. Section 9.0 instructs contractors to keep work areas as quiet as possible when the work is proximal to buildings within which class activities are occurring. The UB SUNY project manager or the EH&S can temporarily halt work activities that cause noise at a level that affects classrooms activities.

The Common Council of the City of Buffalo has issued an ordinance to regulate the level of noise within the city under Chapter 293, Noise, August 13, 1990. Section 293-4 states the following regarding construction:

I. Construction activity.

(1) The performance or engagement in construction work, building, excavating, hoisting, grading, demolishing, dredging or pneumatic hammering within the limits of the city between the hours of 9:00 p.m. and 7:00 a.m. that causes sound which annoys or disturbs a reasonable person of normal sensitivities in a residential real property zone, except for emergency work of public service utilities or as otherwise provided in Subsection I(b) herein [is prohibited].

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BMRC Decommissioning Environmental Report Revision 0 4.8 Cultural and Historical Resources The Buffalo Preservation Board maintains a website with detailed information on various historic districts within the City of Buffalo. The Board is an agency of the City of Buffalo and is administered by the City of Buffalo Office of Strategic Planning. Nine historic districts are listed as either Certified Local Preservation District or National Register Historic Districts. Maps for each of the districts are available for viewing on the website. After viewing the maps, it was determined that the UB South Campus is not located in any of the historic districts. The nearest historic district was the Hamlin Park Local Preservation District.

The historic district known as Hamlin Park is located approximately two miles south-southwest of the BMRC. Hamlin Park developed during the first 25 years of the 19th century as a textbook zone of emergence. The name of the area is derived from a former horse racing park, Hamlins Driving Park, which opened in 1858. The owner of the track, Cicero Hamlin, once touted a race as the Kentucky Derby of the North. Following Cicero Hamlin's death in 1905, Hamlin's racing park lay dormant. After several fires and years of neglect, this three hundred acre parcel of scruffy and largely vacant land lay empty and unused until 1912, when it was bought by a Toronto real estate developer. Homes were developed for the middle class and quickly sold. By the late 1950's and early 60's, Hamlin Park had changed. A transition had occurred as middle-class African-Americans moved in and in the process created the best and the nicest African-American neighborhood in Buffalo. In 1999, the area was officially designated as a Historic District.

Several buildings of historical significance are located within the city limits of Buffalo. A select few worthy of mention include Buffalos City Hall which opened in 1931; St. Pauls Episcopal Cathedral, originally built in the mid-19th century and re-built 40 years later after a fire; and the Theodore Roosevelt National Historic Site, the site of the emergency swearing in as president of Theodore Roosevelt following the assassination of President William McKinley in Buffalo in 1901. The Buffalo Preservation Board lists 89 local Buffalo Landmarks, and 77 state and regional registers within the Buffalo city limits.

There were no sites listed that correlated with addresses or buildings located on or near the South Campus The UB South Campus, also known as the Main Street campus, is the former grounds of the Erie County Almshouse and Insane Asylum, of which four buildings remain (Hayes Hall, the former insane asylum; Wende Hall, a former maternity hospital; Hayes D; and Townsend Hall, a former nurses' quarters). The four buildings were constructed during the mid to late 1880s. The college was designed by architect E.B.

Green in 1910 and was originally intended to resemble Trinity College in Dublin, Ireland. Although these 22

BMRC Decommissioning Environmental Report Revision 0 structures are of a sufficient age and historical significance, they are not listed on any documents for local, state, or national historical recognition.

4.9 Visual/Scenic Resources The BMRC is located in the southeastern section of the UB South Campus. The campus is within an area of heavy residential neighborhood to the south, west, and north. The Veterans Administration Hospital is to the southeast and a golf course is located to the northeast. Some commercial development is present north of the South Campus.

The BMRC is located within what would be classified as a foreground-middle ground zone by the BLM Visual Resource Inventory and Evaluation System. Local residents would rarely glimpse the structure as it is positioned such that the view from the south, southwest, and southeast is obscured by other campus buildings. The population most likely to view the BMRC on a regular basis is the student population. It is unlikely that most students would feel a negative aesthetic impact during the demolition of the BMRC or after the structure is removed.

4.10 Socioeconomic Buffalo is the second most populous city in New York State with an estimated 261,310 residents (2010 Census). The city has seen a steady decline in population numbers since peaking at around 580,132 residents in 1950. The decline has come about because of the loss of industrial base but researchers of the population indicate that the population decline may have finally stabilized. Based on the 2005-2007 American Community Survey Estimates, the population was 53% white, 41% black or African American, 1% American Indian and Alaska Native, 2% Asian, 4.5% of another race and 2.5% from two or more races. Buffalo has sizeable populations of Irish, Italian, Polish, German, Jewish, Greek, Arab, African American, Indian, and Puerto Rican descent.

The nearest private residences that are not under UB control are approximately 450 feet south of the BMRC. The nearest private residences in the direction of the prevailing southwest to northeast winds that are not under UB control are approximately 3,700 feet northeast of the BMRC. There are no heavy industrial facilities near the South Campus. The nearest airport, Buffalo Niagara International Airport is approximately 4.5 miles east-southeast of the campus. The intersection of Interstates 290 and 90, major highways in the area, are approximately 2.5 miles east of the campus. An Amtrak passenger station is located approximately 5 miles south-southwest of the BMRC. Freight by rail is offered by CSX, Norfolk Southern, Canadian Pacific, and Canadian National in the Buffalo area. The United States Coast Guard 23

BMRC Decommissioning Environmental Report Revision 0 Auxiliary, Flotilla 3-3, 9th District, Eastern Region is located 6.25 miles southwest of the campus. There are no other major military bases or facilities in the Buffalo region.

A May 1983 Safety Evaluation Report related to the renewal of the operating license for the Research Reactor at the State University of New York at Buffalo drafted by the United States Nuclear Regulatory Commission was reviewed as part of this report. The Report indicated that no heavy industry, heavy air or ground traffic would constitute an external threat to the integrity of the reactor facility. The report concluded that there was no significant risk from accidents to the reactor because of activities related to the military, industry, or heavy transportation traffic. Since the drafting of the 1983 report, the major interstates I-90 and I-290 have been added to the region to the east, northeast, and southeast of the BMRC. No other significant changes have occurred since 1983 that would alter the conclusions of the report. The addition of the interstates to the area is not considered a significant risk to the reactor at the BMRC.

4.11 Public and Occupational Health Public and Occupational Health activities associated with the current operations of the BMRC include the following:

Containment Building/Reactor All of the spent fuel was shipped to the Department of Energy (DOE) in September 2005 and the unused fuel was sent to North Carolina State University in July 1998. Radiation exposures from the former reactor core are reduced to acceptable levels by water and concrete shielding. The Primary Coolant heat exchanger was removed and disposed of as radioactive waste in 1994. The containment building had a significant amount of legacy waste on all three levels until the Pre-decommissioning Cleanout project in the summer of 2011. All of the reactor components remain in the pool and the components from the 1991 reactor rebuild were stored under shielding on the Neutron Deck of the building until they were sent for burial at the EnergySolutions facility in Clive, UT during the Pre-decommissioning Cleanout Project.

There are no regularly scheduled activities within the containment building.

ENERCON conducted a Historical Site Assessment (HSA), and released a report with findings on March 17, 2010. The HSA reported that events and operations have transpired that have impacted the containment building. The roof, interior and all the systems and structures in the building are designated as Impacted Areas. These areas and/or components have been designated as impacted due to elevated levels of residual radioactive materials or the potential for the presence of residual radioactive materials and are expected to require, in part or whole, remediation and/or disposal at an NRC licensed disposal 24

BMRC Decommissioning Environmental Report Revision 0 facility. The HSA reported that the primary contaminants of concern for the entire BMRC site were H3, C14, Mn54, Co58, Co60, Eu152, Eu154, Ag108m, Ag110m, Sb124, La140, and Ni63. Since the plant has been shutdown since June 23, 1994, the Co58, Sb124, Ag110m and the La140 have decayed significantly due to their short half-life. The HSA concluded with a recommendation that scoping/characterization surveys be conducted for the containment building.

ENERCON conducted a site characterization study and released a Site Characterization Report (SCR), on May 30, 2011. The study did not include the activation of reactor components and the pool liner. The SCR offered the following findings:

x Control Deck:

o Many areas were being used to store legacy radioactive wastes. The elevated background from the radioactive material storage areas made it difficult to determine the presence of residual radioactive materials on building surfaces in these areas. However, the limited radiological measurements collected in the areas of the Control Deck did not identify removal or fixed radiological activity. Therefore, it seems reasonable to conclude that the building surfaces in these areas are not contaminated. Additional testing for confirmation of this conclusion will be performed when all legacy radioactive wastes are removed from the area.

o Room 202 was the location of a strontium-90 (Sr-90) release that contaminated areas of the walls adjacent to the California hood. Sections of Plexiglas have been affixed to the walls to afford temporary shielding from the strontium.

x Gamma Deck o Most rooms on the Gamma Deck are currently being used to store radioactive waste, making characterization of building surfaces difficult. Additional testing of building surfaces will be performed subsequent to the removal of all items from the area. All items inside the Hot Cell, including the Hot Cell, should be considered radioactive waste.

x Neutron Deck o The Neutron Deck is the area where the majority of radioactive waste materials awaiting disposal are currently stored. Subsequent to the removal of all radioactive material, a more detailed investigation of the building surfaces will be performed.

o Concrete cores of the floor were collected at six locations. Coring confirmed that the concrete was originally poured directly onto bedrock. Because a sand or gravel layer was not placed on top of the bedrock prior to pouring of the concrete, no preferential pathway exists between the Neutron Deck and the bedrock. The analytical results of the testing indicated background levels for the evaluated radionuclides.

o Concrete cores of the bioshield were collected at six locations. The data from the testing indicates measureable radioactivity due to neutron activation in certain areas exceeds 12 inches in depth.

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BMRC Decommissioning Environmental Report Revision 0 Administrative/Laboratory Wing The administrative wing also has radioactive materials stored in various areas. In addition, the UB radiological instrument calibration facility is located in the Laboratory Wing. The administrative wing is used infrequently for two purposes. One purpose is an ion chamber calibration facility that is located in the basement of the administrative wing. Secondly, the top floor has asbestos-containing floor tile in some areas. The UB Maintenance Department uses this floor to train maintenance and custodial personnel how to safely strip and wax this type of floor tile.

Based on the HSA, dated March 17, 2010, events and operations have transpired that have impacted the administrative wing. The roof, interior and portions of the systems and structures in the building are designated as Impacted Areas. These areas and/or components have been designated as Impacted due to elevated levels of residual radioactive materials or the potential for the presence of residual radioactive materials and limited portions are expected to require remediation and/or disposal at an NRC licensed disposal facility.

The ENERCON SCR offered the following findings:

x Control Deck o Rooms 208, 208A, 209, 209A, 209B, 210, 212, 213, 214, 215, and 216 showed no elevated measurements of residual radioactive materials.

x Gamma Deck o Rooms 110, 111, and 113 showed no elevated measurements of residual radioactive materials.

o Rooms 114, 115, 115A, 117, and 118 contain radioactive materials. These spaces may receive additional surveys subsequent to the removal of radioactive materials.

x Sub-Basement o Room NO3 contains a 1,000-gallon tank containing six inches of sludge with elevated levels of radioactive materials.

o Room N16 contains a 10,000-gallon tank internally contaminated with radioactive materials.

Exterior Sub-Surface x Soil samples characterized for radionuclides outside the facility indicated that remediation of soil will not be necessary.

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BMRC Decommissioning Environmental Report Revision 0 Tank Farm x The laboratory analysis of samples collected from each of the five tanks indicated that elevated levels of radionuclides are present in the waste tanks. The liquids will require proper handling to ensure radioactive materials are not released to the environment. The water volume in each tank is 75-100% of their maximum capacity.

x The tank farm has sample tubes that extend down outside of the tank to the bottom of the enclosure and the bottom of the concrete berm. These sample tubes, known as "tell-tale" samples, have not indicated any major leaks from the tanks.

N 16 Tank Vault x Testing of dirt and debris on the bedrock in the N16 Tank room indicated impacts greater than the NRC and New York State (NYS) screening values for Co-60. Ag-108m was elevated and has a high potential to be greater than the DCGLs stated in the decommissioning plan.

1,000 Gallon Tank x Water and sediments were collected for characterization from the 1,000-gallon tank located in Room N03. The tank sediments contained elevated levels of the radionuclides Co-60, Ag-108m, Ni-63, and Sr-90. The water sample contained Ni-63.

Extraneous Sources Sources of radiation within the BMRC were associated primarily with legacy wastes stored in a variety of areas in both the administrative wing and the containment Building. The reactor is inoperable and all fuel and spent fuel have been shipped from the site. There are currently no regularly scheduled activities within the BMRC except for occasional training and routine inspection. UB has developed written policies that ensure that authorized persons only are allowed into the BMRC. An intrusion alarm system has been installed and is operational any time the BMRC is unoccupied. The alarm system is checked on a weekly basis.

UB has developed a written monthly and quarterly check sheet to ensure that failsafe systems are operational within the BMRC. Systems inspected and calibrated include, but are not necessarily limited to, the Building Gas electronics, Building Particulate electronics, the Control Deck Continuous Air Monitor (CAM), and the emergency damper closure time in response to a manual damper scram.

Groundwater samples are collected from the underground wastewater vaults on a monthly basis to test for radionuclides.

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BMRC Decommissioning Environmental Report Revision 0 Routine Monitoring According to UB EH&S staff, approximately ten to twelve UB employees are required to wear personal Thermoluminescent dosimeters (TLD) during activities in the BMRC. The personnel badges are analyzed on a quarterly basis. The EH&S staff report that personal exposures are less than the minimum reportable dose of 10 mrem per sampling period. TLDs are also placed at various interior and exterior locations to monitor the general areas of the BMRC. The TLDs are submitted to a testing laboratory for analysis on a monthly basis. The exterior TLDs read 0 to 4 mrem above background levels per reporting period. The analytical results of the interior TLDs typically range from 0 to 2 mrem per month, however, the TLD positioned on the bridge over the reactor pool yields the highest readings of 3 to 7 mrem per period.

A 1963 report on Environmental Monitoring Program for Western New York Nuclear Research Center Inc. and UB was available for review. The report detailed environmental monitoring of the former BMRC and surrounding area beginning on March 1, 1961. Sampling for radioactivity was performed on surface water, soil, vegetation, air, and atmospheric fallout. The report concluded that activities at the nuclear research facility had not caused any detectable increase in the radiation background of the environment.

4.12 Waste Management A Waste Management Plan (WMP) will be implemented for the disposal of the waste generated during the BMRC D&D project. The WMP will include detailed guidance for the characterization, sampling, classification, segregation, handling, packaging, manifesting, transporting and disposal of all waste categories.

ENERCON performed a waste study for the BMRC of the legacy waste items and released a report dated December 2009, Study to Determine Quantities, Types, and Disposition of Waste at the Buffalo Materials Research Center. The waste study divided all of the materials that will eventually require disposal into five potential treatment categories. These are as follows:

x Items for on-site survey and unrestricted release (free release) x Bulk Survey for Release (BSFR) x Off-site Processing x Lead Processing x Direct Disposal The waste study discusses the five potential treatment categories in depth and the reader of this report is referred to the waste study for additional detail, if needed. The legacy waste was removed during the Pre-Decommissioning Cleanout. The R-77 license authorizes the licensee to ship radioactive material.

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BMRC Decommissioning Environmental Report Revision 0 5.0 ENVIRONMENTAL IMPACTS The following sections describe the potential environmental impacts that may result from the decommissioning activities of the BMRC.

5.1 Land Use Impacts Land use impacts are expected to be significant during the decommissioning activities. The construction activities involved during decommissioning will require the removal of all radioactive materials from the BMRC, dismantling the reactor and any peripheral support systems, removal of UST and associated piping, removal of materials associated with the previously demolished cooling tower, demolition of both wings and removal of any waste.

Following the decommissioning, the site will be returned to usable space by backfilling the void created to match the normal grade of the surrounding areas. Any further land use impacts are expected to be minimal.

5.2 Transportation Impacts The primary project impacts to the environment from transportation would occur when various forms and quantities of radioactive, nonradioactive, and hazardous wastes are shipped from the BMRC facility during decommissioning activities. During transport, all materials are to be effectively packaged to prevent significant radiation or other impacts external to the truck. Thus, the primary impacts are accident risk and emissions/noise from the trucks themselves.

The dose consequence from transportation accidents occurring with radioactive waste could be higher than the contamination accident scenarios described in Section 5.6 and Section 5.12.2 because high-activity reactor components could be involved. As such, there is a potential for a moderate dose consequence of between one and 25 mrem for the public following a transportation accident. However, adherence to NRC and DOT radioactive material packaging and transportation requirements is considered a sufficient control measure for mitigating transportation-related incidents.

Local streets will be utilized initially to transport radioactive waste and construction debris associated with the decommissioning and demolition of the BMRC. Interstates 90 and 190 are less than five miles from the site and will likely be utilized as transport routes to the eventual disposal destinations. Therefore, it is expected that there will be minimal transportation impacts due to increased trucking activity during the decommissioning activities.

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BMRC Decommissioning Environmental Report Revision 0 5.3 Geology and Soils Impacts Based on the Site Characterization results three (3) samples collected from the N16 room in the sub-basement indicated impacts greater than the NRC Screening values for Co-60. Additionally, Ag-108m was elevated and has a high potential to be greater than the DCGL developed as part of the Decommissioning Plan. The impacted dirt and debris will be removed during the decommissioning activities. Therefore, it is expected that there will be no geology or soil impacts as a result of decommissioning of the BMRC.

5.4 Water Resources Impacts ENERCON anticipates that there will be no impact to water resources as a result of decommissioning of the BMRC. All water generated as a result of the decommissioning of the BMRC will be contained and disposed of in accordance with all UB, federal, state and local regulations. No water other than water mist for dust control will be introduced to the subsurface environment or into the natural environment outside of the facility. No water will be discharged to the stormwater system, unless it has been sampled to verify it is within regulatory limits.

5.5 Ecological Resource Impacts No ecological resource impacts are expected, as all work will be performed within the BMRC. Potential lay down areas for equipment will only be placed in developed areas. Decommissioning the reactor would have no impact on amphibians, reptiles, birds, fish, mammals or threatened species.

5.6 Air Quality Impacts The decommissioning activities could minimally impact the air quality near the BMRC from both stationary and mobile source emissions.

A small increase in mobile source emissions, such as carbon monoxide and nitrogen oxides, could be released from contractors trucks and cars. Due to the temporary nature of the decommissioning activities, effects from mobile source emissions would be low.

Stationary emissions from the decommissioning activities could occur during demolition and soil remediation activities. The emissions from the building demolition, subsurface material excavations and general construction work are expected to be minimal. All applicable dust control measures will be implemented as part of the decommissioning construction activities. Standard asbestos abatement procedures implemented by a contractor licensed by the State of New York will be used to remove any 30

BMRC Decommissioning Environmental Report Revision 0 asbestos-containing materials. Site workers would be protected during decommissioning activities through air monitoring and the use of air purifying respirators when required.

An uncontrolled release of airborne radioactivity could occur during cutting and demolition activities involving activated materials. Such activities may take place inside temporary containment structures equipped with high-efficiency particulate air (HEPA) filter ventilation systems or other systems, if applicable. The failure of a containment structure could result in the release of airborne radioactive materials into environment. Therefore, an alarming CAM will be used in the work areas to warn against the release of airborne radioactivity.

While the actual concentrations of airborne radioactive materials are unknown at this time, the dose consequence of an uncontrolled release is expected to be low (<1 mrem off-site impact and < 25 mrem to on-site workers). As such, safety management operations (standard engineering and administrative controls) are sufficient for protection against such accidents.

5.7 Noise Impacts Noise impacts during the decommissioning of the BMRC are expected to be those associated with normal construction activities. The UB project manager or the EH&S can temporarily halt work activities that cause noise at a level that affects classrooms activities. The construction activities will take place between 7:00 am and 9:00 pm to ensure compliance with the City of Buffalo noise regulations.

5.8 Historical and Cultural Resources Impacts The decommissioning and demolition of the reactor and support facilities is not expected to have any historical or cultural impacts.

5.9 Visual/Scenic Resource Impacts Visual/Scenic Resource impacts associated with decommissioning are primarily visual and relate to the structures and visual attributes of the decommissioning site. The impact of decommissioning on site aesthetics (e.g., truck traffic, demolition, noise) is limited in duration. Therefore, any impacts are temporary and are expected to be minimal.

5.10 Socioeconomic Impacts No socioeconomic impacts are expected from the decommissioning of the BMRC.

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BMRC Decommissioning Environmental Report Revision 0 5.11 Environmental Justice High and adverse health, economic or environmental effects to local low-income and minority populations characterize environmental justice. There is no reason to believe that low-income or minority populations would be affected by BMRC decommissioning.

5.12 Public and Occupational Health Impacts Nonradiological and radiological impacts from decommissioning activities are expected to be minimal.

The work associated with nonradiological and radiological controls will be conducted within approximately 150 feet of the UB property boundary and within 450 feet of the nearest residential area.

Figure 5, on the next page, identifies the BMRC location in association with the property boundary and the nearest fulltime resident.

The following sections describe the potential nonradiological and radiological impacts associated with the decommissioning activities.

5.12.1 Nonradiological Impacts As part of the decommissioning process nonradiological impacts will be associated with the removal of any general building materials and asbestos in the decommissioning area. The removal of asbestos containing materials will be conducted in accordance with applicable state and federal regulations.

Removal of general building materials and debris will be conducted utilizing industry standard controls and applicable state guidelines.

5.12.2 Radiological Impacts There is a potential for radiological accidents during the BMRC facility decommissioning project resulting from the uncontrolled release of radioactive materials to the work area or the environment.

These releases are most likely associated with the mismanagement of contaminated liquids in the waste tanks. Uncontrolled releases of airborne contamination could also occur during the demolition of the reactor tank and segmentation of the activated bioshield. An uncontrolled release of radioactive material could also occur during a transportation accident.

The proper handling of the waste liquids by draining the tanks prior to moving them is deemed adequate to prevent uncontrolled release. The demolition of the reactor tank and the segmentation of the bioshield should be conducted inside containment with HEPA filtration to help prevent an unplanned release to the environment.

The accidental dropping of an activated reactor component was also considered as a potential accident.

However, because the more highly activated components are located under water and the water is 32

BMRC Decommissioning Environmental Report Revision 0 purified, the surface contamination on these parts is minimal and would not release significant quantities of radioactive materials during such an incident. Such an incident would most likely result in additional external exposures. The airborne release fraction from dropped metal or soil removal is relatively low.

Very minimal amounts of soil are expected to require remediation and removal from the site.

A fire is another possible source of an uncontrolled release of radioactive materials. However, the majority of the combustibles that will be present onsite will be clean materials. Potentially contaminated combustibles may include dry active waste such as personal protective clothing, rags and towels used for site cleanup and decontamination. The radioactivity contained in these materials would not be high enough to result in a significant release during such an incident. There will be no fissile materials located on the site that could result in a criticality incident because all nuclear fuel has been removed.

5.13 Waste Management Impacts The BMRC decommissioning will generate soil LLRW, mixed waste (i.e., contaminated lead and cadmium), hazardous waste (i.e. ACM, oils and fluid drained from equipment). These wastes will be handled, stored and disposed of according to applicable state and federal regulations. The DOC will coordinate with the waste disposal site(s) regarding the sites waste acceptance criteria and pre-shipment processing requirements.

Waste processing may include volume reduction through compaction or segmentation, neutralization, stabilization or solidification. Due to the limited size of the facility and work area, concrete rubblization beyond that required for demolition is not expected to occur on-site. Complying with written procedures, standard work practices and operating with the limits of the NRC license will ensure safe waste processing. The decisions as to the type and degree of waste processing will primarily be based on economics that weigh the costs of additional handling and processing compared to transferring the material off-site for treatment and/or disposal.

After the characterization surveys and sampling are complete, wastes will be wrapped, bagged and/or containerized and staged in the appropriate designated area. Items and containers will be properly labeled as Radioactive Material and the label will indicate the external dose rate from the material. Radioactive wastes will be stored in properly secured radioactive materials storage areas. Waste material logs will be maintained for materials placed in disposal and shipping containers.

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BMRC Decommissioning Environmental Report Revision 0 Figure 5: Decommissioning Boundary 34

BMRC Decommissioning Environmental Report Revision 0 5.13.1 Radiological Impacts Prior to disposal, all waste streams will be properly characterized according to the requirements of the disposal facility. This characterization will include quantification of primary radionuclides of concern as well as hard-to-detect radionuclides. Additionally, those radionuclides that have specific limits for Class A waste will be directly quantified or estimated based on ratios to concentrations of other radionuclides.

All waste will be shipped to an acceptable waste disposal site in accordance with applicable NRC, DOT, and NYS regulations regarding waste packaging, labeling and placarding. Each LLRW shipment will be accompanied by a shipping manifest as specified in Section I of Appendix F to 10 CFR 20, Requirements for Low-Level Waste Transfer for Disposal at Land Facilities and Manifests. The waste will be manifested consistent with its classification. Only licensed transporters will be used to transport wastes from the BMRC facility.

Mixed wastes may be shipped to an NRC licensed processing facility or directly to an NRC licensed land disposal facility depending on the nature of the waste and the treatment options available.

The consequence levels discussed in this section are described in more detail in the DOE Standard DOE-STD-1120-2005, Integration of Environment, Safety and Health Into Facility Disposition Activities (DOE 2005) 35

BMRC Decommissioning Environmental Report Revision 0 6.0 MITIGATION MEASURES The following table identifies the areas that may be impacted by the decommissioning of the BMRC and the mitigation measures:

x Land Use x Transportation x Geology and Soils x Meteorology, Climatology and Air Quality x Noise x Public and Occupational Health x Waste Management Mitigation measures that will be used to reduce potential impacts in these areas will include:

Affected Environment Mitigation Measures Land Use Following strict site demolition and restoration procedures.

Transportation Proper loading and shipping of containers and/or materials going off-site.

Geology and Soils Proper removal and disposal of soils impacted above regulatory levels.

Meteorology, Climatology and Air Quality Build negative air pressure tents (when feasible) around work areas so that potential air contaminants are contained within work areas.

Using water sprays to minimize the amount of dust generated. Following applicable construction and abatement procedures.

Noise Limiting the time of increased noises to daylight hours, with a schedule that takes account of the UB academic calendar.

Public and Occupational Health Constant radiological and air monitoring during decommissioning activities will ensure that the public and the occupational workers are safe.

Waste Management Proper decontamination and segregation of waste will ensure that wastes are disposed of properly.

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BMRC Decommissioning Environmental Report Revision 0 7.0 ENVIRONMENTAL MONITORING AND MEASUREMENT PROGRAMS This section describes the environmental monitoring and measurement programs that will be used during the decommissioning of the BMRC.

7.1 Radiological Monitoring Frequent radiological monitoring will be performed during the BMRC decommissioning activities. CAMs will be used to monitor the air during selected decommissioning activities. Survey stations will be set up to monitor personnel and equipment. All packaged waste materials will be surveyed prior to leaving the facility for disposal.

7.2 Physiochemical Monitoring Monitoring for chemical/hazardous constituents such as asbestos, lead, cadmium and mercury will be performed as needed during the BMRC decommissioning activities. A thorough asbestos inspection will be performed prior to decommissioning activities and all materials suspect to contain asbestos will be characterized. Asbestos-containing materials will be removed by a trained and licensed New York Department of Labor Asbestos Abatement Contractor prior to the eventual demolition of the facility.

7.3 Ecological Monitoring No ecological monitoring is required during the decommissioning of the BMRC.

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BMRC Decommissioning Environmental Report Revision 0 8.0 COST BENEFIT ANALYSIS Detailed cost estimates for the completion of the BMRC decommissioning activities are included in the Decommissioning Plan. The following are some of the benefits associated with completion of the decommissioning of the BMRC.

x Increased public health and safety; x Decreased operating and maintenance costs; x Free release the site for reuse; x Termination of the NRC licenses.

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BMRC Decommissioning Environmental Report Revision 0 9.0

SUMMARY

OF ENVIRONMENTAL CONSEQUENCES The following is a summary of the environmental consequences related to the implementation of the proposed action:

Affected Environment Environmental Consequences Short Term Impacts - Construction and demolition of the BMRC buildings, excavation and removal Land Use of all associated tanks and piping and removal of all subsurface materials. Following demolition conducting site restoration to usable space.

Transportation Short Term Impacts - Hazardous, solid, and radioactive waste transportation Soils will be impacted by removal of structures Geology and Soils and tanks. Fill will be introduced to the site as necessary and the area will be restored to grade.

Water Resources No Impacts Ecological Resources No Impacts Meteorology, Climatology, and Air Quality Short Term Impacts - Possible air quality issues associated with construction activities.

Noise Short Term Impact - Noise from heavy equipment and demolition of the structures.

Cultural and Historic Resources No Impacts Socioeconomic No Impacts Short Term Impact - Possible radiological Public and Occupational Health exposure, possible asbestos fiber exposure, and dust above background.

Waste Management Short Term Impact - Disposal of solid, hazardous, and radioactive waste.

There will be no adverse or long-term environmental impacts as a result of the implementation of the proposed action.

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BMRC Decommissioning Environmental Report Revision 0

10.0 REFERENCES

1. NUREG-1748, Environmental Review Guidance for Licensing Actions Associated with NMSS Programs (NRC 2003b).

2. 10CFR20, Standards for the protection against Radiation

3. 2006 New York State Low-Level Radioactive Waste Transportation Report, March 2008, New York State Department of Environmental Conservation Division of Solid & Hazardous Materials

4. 2035 Long Range Transportation Plan Update, May 2010, Greater Buffalo-Niagara Regional Transportation Council

5. BMRC Historical Site Assessment, Rev. 0, March 17 2010, ENERCON

6. BMRC Study to Determine Quantities, Types, and Disposition of Waste at the Buffalo Materials Research Center, December 2009, ENERCON

7. BMRC Historical Site Assessment, Rev. 0, March 17 2010, ENERCON

8. BMRC Characterization Plan, November 30, 2010, ENERCON

9. Chapter 293, Noise, City of Buffalo City Ordinances, August 13, 1990, Common Council of the City of Buffalo

10. Custom Soil Resource Report for Erie County, New York, BMRC at SUNY, May 2, 2011, United States Department of Agriculture Natural Resources Conservation Service

11. Issuance of Amendment No. 26, To Facility Operating License No. R University of New York Buffalo [Buffalo Materials Research Center (BMRC) Research Reactor] (TAD No. MC3050), May 4, 2005, NRC

12. Landmarks in Buffalo, New York, Buffalo as an Architectural Museum,

<http:buffaloah.com/a/landmks/landmks6.html>, May 5, 2011, Landmarks and Historic Districts in Buffalo

13. Niagara Frontier Urban Area Freight Transportation Study, Technical Memorandum No. 2, Freight Transportation System Profiles, Project No. 06 Freight, February 20, 2008, Wilbur Smith Associates

14. NUREG-0982, Safety Evaluation Report related to the renewal of the operating license for the Research Reactor at the State University of New York at Buffalo, May 1983, NRC 40

BMRC Decommissioning Environmental Report Revision 0

15. NUREG-1748, Environmental Review Guidance for Licensing Actions Associated with NMSS Programs, Final Report, August 2003, NRC

16. Report On Environmental Monitoring Program For Western New York Nuclear Research Center Inc., and the University of Buffalo, 1963, The University of Buffalo and the Western New York Nuclear Research Center, Inc.

17. Specifications for Nuclear Research Center for the University of Buffalo, Buffalo, New York, Contract No. 1, General Construction, June 6, 1959, Office of J. Fruchtbaum

18. University at Buffalo R-77, Operating Procedure #59, Operation of the Intrusion Alarm System, Revision Date January 9, 2009, UB EH&S

19. University at Buffalo R-77, Operating Procedure #77, Quarterly Checks, Revision Date January 9, 2009, UB EH&S

20. Visual Resource Management, BLM Manual 8400, US Department of Interior, Washington, D.C.,

April <http:www.blm.gov:80nstc/VRM/8400.html>, May 3, 2011, Bureau of Land Management 41