ML12025A125

Ford Nuclear Reactor - Technical Specification Amendment Request Decommissioning Plan - Revised Section 4.0 (Final Status Survey)

ML12025A125
Person / Time
Site:	University of Michigan
Issue date:	01/20/2012
From:	Driscoll M University of Michigan
To:	Document Control Desk, NRC/FSME
References
Download: ML12025A125 (66)

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<O$OTEH University of Michigan O0SEH Occupational Safety &

Occupational Safety & Environmental Health Campus Safety Services Building 1239 Kipke Drive, Ann Arbor, MI 48109-1010 SEnvironmental Health Phone: 734 647-1143

• Fax: 734 763-1185 January 20, 2012 Document Control Desk U.S. Nuclear Regulatory Commission Two White Flint North 11545 Rockville Pike (Mail Code: 03H8)

Rockville, Maryland 20852-2738 RE: Ford Nuclear Reactor - Technical Specification Amendment Request Decommissioning Plan - Revised Section 4.0 (Final Status Survey)

Docket 50-2 / License R-28 Licensing Branch:

The University of Michigan (U-M) is requesting an amendment to the Ford Nuclear Reactor (FNR) Decommissioning Plan that was submitted to the Commission in a correspondence dated January 10, 2006. As previously requested in an amendment request dated April 8, 2011, the U-M requests Section 4.0 ('Final Status Survey') of the Decommissioning Plan be revised in its entirety to more accurately reflect the present condition of the facility. A copy of the most recently revised Final Status Survey (FSS) Plan is enclosed with this correspondence.

It should be noted that several sections and Table 4-2 of the enclosed FSS Plan have been revised since the April 8, 2011 submittal. As requested by the NRC, Table 4-2 now lists the references used for classifying the listed survey units in accordance with NUREG-1 575 (MARSSIM).

This FSS Plan incorporates project-specific information relative to post-remediation facility conditions and potential radiological contaminants, guidelines for residual building surface and soil contamination levels, sampling and measurement methods, survey unit identification and classification, and data evaluation techniques. A Quality Assurance Project Plan (QAPP),

applicable to FSS activities, has also been prepared.

The U-M is requesting that three survey units associated with the FNR decommissioning project be incorporated by license amendment into the FSS Plan at a later date. These three survey units include: (1) the below-grade foundation drain tile system located outside the FNR on the east side of the building, (2) the FNR foundation cavity located beneath the FNR southwest freight door, and (3) the below-grade storage ports (excavated & removed) located on the west side of the building.

These three survey units will continue to be assessed, surveyed, and evaluated outside the bounds of the enclosed FSS Plan, but following all of the protocols specified in the FSS Plan.

Documentation for each of the three units will be submitted at a future date as amendments to the enclosed FSS Plan.

2 Refer to the two sketches enclosed with this correspondence depicting the foundation drain tile system adjacent to the FNR basement and the FNR foundation cavity located on the first floor.

In addition, please note that the RESRAD computer model code has been used to develop DCGLs for the small volume of sub-surface soil in the vicinity of the foundation drain tile, the floor cavity, and the former storage port area (Appendix E of the FSS Plan).

While the remediation effort and radiological assessment of these three isolated locations continue, the U-M would like to proceed with the Final Status Survey within the general FNR facility.

Thank you for your time, effort, and consideration with respect to this FNR license (R-28) amendment. Please do not hesitate to contact me at OSEH / Radiation Safety Service [(734) 647-2251] should you have any questions or comments regarding the revised Final Status Survey Plan. We look forward to your approval so we can initiate the final status survey of the FNR facility.

Sincerely, Mark L. Driscoll Director / Radiation Safety Officer Radiation Safety Service / OSEH MLD/TGA/mId NRCFNRR-28AmendmentD&DFSSPlanSec4011912.doc cc: Terry Alexander, Executive Director, OCS Mark Banaszak Holl, Associate Vice President, OVPR Robert Blackburn, Manager, Laboratory Operations, MMPP Theodore Smith, FNR Project Manager, NRC Headquarters (Mailstop T-8F5)

Jeremy Tapp, Health Physicist, NRC Region III FNR Decommissioning File

March 2011 FORD NUCLEAR REACTOR BASEMENT or - I r f0CA'DAI0C,,pI ';)'*RA7p' ILE sisvem (41,ýDs;ffivmp- 00 20 Fdiat Exposed soil Figure 4.1 Floor Plan of the Ford Nuclear Reactor Building Basement 5

March 2011 FORD NUCLEAR REACTOR FIRST FLOOR ElJ L4L Open 4101 0 10 20 Feet E Expose d Soil Figure 4-2 Floor Plan of the Ford Nuclear Reactor First Floor 6

January 2012 4.0 Final Status Survey Plan 4.1 Introduction The Ford Nuclear Reactor (FNR) at the University of Michigan (UM) was a light-water cooled and moderated open-pool design reactor. The reactor was licensed (License R-28, Docket 50-2) by the US Nuclear Regulatory Commission (NRC to operate at a power level of 2 Megawatt (MW) thermal. The reactor began operation in 1957 and provided neutron and gamma irradiation services, neutron beam port experimental facilities, and training facilities for use by faculty, students, and researchers from the UM, other universities, and industrial organizations. In July 2003, the reactor was shut down. Fuel was removed from the facility in December 2003, followed closely by initiation of decommissioning activities. These decommissioning activities are described in University of Michigan Decommissioning Planfor the Ford Nuclear Reactor, Revision 1, January 5, 2006 '(Decommissioning Plan). The objective of the decommissioning is to remove radiological materials and equipment associated with the FNR-licensed operations, such that radiological conditions satisfy NRC criteria for future unrestricted use of the facility and thus permit termination of the NRC license. A Final Status Survey (FSS) will be performed to demonstrate that these NRC criteria have been satisfied. This document describes the FSS Plan.

This Plan replaces the "Proposed Final Status Survey Plan" (Chapter 4.0) of the January 2006 Decommissioning Plan. It was prepared in accordance with the guidelines and recommendations presented in NUREG-1757, ConsolidatedNMSS Decommissioning Guidance, and NUREG-1575, Multi-Agency Radiation Survey and Site InvestigationManual (MARSSIM). The process emphasizes the use of Data Quality Objectives (DQOs) and Data Quality Assessment (DQA), along with a quality assurance/quality control program. The graded approach concept will be followed to assure that survey efforts are maximized in those areas having the greatest potential for residual contamination or the highest potential for adverse impacts of residual contamination.

This Plan incorporates project-specific information, relative to post-remediation facility conditions and potential radiological contaminants, guidelines for residual building surface and soil contamination levels, sampling and measurement methods, survey unit identification and classification, and data evaluation techniques. A Quality Assurance Project Plan, (QAPjP), applicable to the FSS activities, has also been prepared.

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January 2012 4.2 Facility Description The Ford Nuclear Reactor is located at 2301 Bonisteel Boulevard on the North Campus of the University of Michigan, approximately 1.3 miles northeast of the central business district of Ann Arbor, Michigan. The FNR building is a windowless, reinforced concrete structure with brick veneer. Internal walls are of concrete block. The footprint of the building is approximately 70 feet (21.3 m) long and 68 feet (20.7 m) wide; the building height is approximately 69 feet (21.0 m) of which about 14 feet (4.3 m) and 23 feet (7.0 m) are below grade on the east and west sides, respectively. During operation, the facility consisted of four levels - reactor access and control (3rd floor), maintenance and other support facilities and systems (2nd floor), beamport experimental area (1st floor), and liquid cooling and waste systems (basement). There was a cooling tower above the reactor pool level. FNR is contiguous with the Michigan Memorial Phoenix Project (MMPP). Some systems, including exhaust ventilation ducts from the beamport floor and neutron activation laboratory, piping to liquid waste retention tanks, access ports to the MMPP hot cells, and neutron activation pneumatic transfer systems were shared by the two buildings. The portions of piping, leading to liquid waste retention tanks in MMPP will be retained and used under the UM broad scope NRC license.

The FNR utilized low-enriched uranium Material Test Reactor heterogeneous plate-type fuel. The reactor core was suspended about 20 feet below the surface of the 10 feet x 20 feet x 27 feet deep pool, containing approximately 5.0 x 104 gallons of deionized water. The pool was lined with ceramic tile and surrounded by a biological shield of barytes concrete. Spent fuel, reactor handling tools, and miscellaneous experimental equipment were stored in the pool. Pool water was purified by a deionizer system. The primary cooling system was a closed loop and included a heat exchanger and associated piping. The secondary cooling system was a counter-flow heat exchanger, and heat was dissipated to the atmosphere through an evaporative cooling tower on the building roof. The deionizer and heat exchanger systems were located on the basement level.

Treated water and water from seepage and leaks was collected and pumped to retention tanks in the adjacent Michigan Memorial Phoenix Project.

General reactor building exhaust was through the FNR facility stack. Localized exhausts for the experimental areas, source storage ports, and laboratory hoods were filtered through HEPA units and discharged through the MMPP ventilation systems. The FNR was serviced by sanitary and storm liquid waste systems, but potentially contaminated liquids were not discharged through these systems.

There is a French drain system external to the foundation of the FNR building.

Fifty storage ports extended through the west wall of the first floor of the facility, into the soil external to the building. These ports were used to store irradiated components and one of the ports housed two PuBe neutron sources, which are licensed on SNM-179. The sources have been relocated to a storage facility in the MMPP facility, where control and oversight are conducted under the University's Radiation Safety Serv ice/ OSEH.

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January 2012 Chapter 2 of the Decommissioning Plan describes the Ford Nuclear Reactor (FNR) facility, its operational history, and the radiological status, prior to remediation and decontamination actions. In addition to the removal and disposition of reactor core assembly and miscellaneous reactor operating materials and tools, remedial actions included extensive removal, disposition and/or decontamination of the following potentially contaminated equipment, components, building areas, and materials:

" Biosafety shield,

" Hold-up tank & pipes,

• Primary pumps & pipes,

" Secondary heat exchanger & pumps,

" Hot & cold DI system,

" Floor drains and drain piping,

" Hot & cold sumps (including overflow pit),

" Transfer chute,

• Thermal column door and trench,

" Underground storage ports (west side of building) and surrounding soil,

" Beam port floor janitor closet tub / drain & water lines,

• Supply & exhaust ventilation system,

• Heating system,

• Electrical (including exposed conduit and most wire),

" Men's & women's rest room (toilets / sinks / water),

• 1st and 2nd floor janitor's closet (water & drains),

" Primary water treatment room,

" Reactor bridge (removed, but saved for historical purposes),

• Control room (saved control console),

" Counting room,

" Pneumatic transfer system and tube room / lab, and

" Fuel vault.

Large sections of the concrete structure with a potential for containing volumetric contamination, due to its location relative to the reactor core, were removed and sampling was performed to demonstrate that volumetric contamination is not present in the remaining concrete. Also, sections of cracked concrete foundation and slab were removed to assure that leakage did not result in contamination of sub-floor soil. Paint and floor/wall coverings have been removed from building surfaces, considered to be potentially contaminated, and the surfaces smoothed to facilitate surveys. Surfaces have been vacuumed and wiped down to remove loose contamination; routine radiological control surveys have demonstrated the absence of removable activity.

Municipal water supply and sanitary, and storm drain systems remain.

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January 2012 Figures 4-1 through 4-5 are floor plans of the FNR facility, illustrating the as-left conditions.

4.3 Radiological Contaminants and Criteria Samples of construction materials and soil were obtained and analyzed during the remediation process and after remediation was completed to identify radiological contaminants that might remain at the time-of the final status survey. The process, whereby the residual potential contaminants was determined, is described in Appendices A and D. Potential contaminants in both soil and structural surfaces are:

• Co-60

• Ag-108m

• Ag-110m

• Cs-137 Future uses of the former FNR facility have not yet been completely defined; however, it is likely that the facility will continue being used in some capacity for University of Michigan academic programs. For these reasons the most restrictive exposure scenarios, i.e., building occupancy and residential farmer are 4

January 2012 FORD NUCLEAR REACTOR BASEMENT SCALE 0 10 20 Feot Exposed Soil Figure 4.1 Floor Plan of the Ford Nuclear Reactor Building Basement 5

January 2012 FORD NUCLEAR REACTOR FIRST FLOOR SCALE 010 20 Feet Exposed.Soil Figure 4-2 Floor Plan of the Ford Nuclear Reactor First Floor 6

January 2012 FORD NUCLEAR REACTOR2N FLOOR SCALE 0 1.0 20 Feet Figure 4-3 Floor Plan of the Ford Nuclear Reactor Second Floor 7

January 2012 FORD NUCLEAR REACTOR THIRD FLOOR MO~ -01 -814 U-I SCALE 0 10 20 Feet Figure 4-4 Floor Plan of the ford Nuclear Reactor Third Floor 8

January 2012 FORD NUCLEAR REACTOR 4TH FLOOR Sar#3.

OpvenBelow F Q

/\

• /

V--"-"-- ------....

• N 0.o

• ~N -

SCALE 0 10 20 Feet Figure 4-5 Floor Plan of the Ford Nuclear Reactor Fourth Floor 9

January 2012 assumed for potential future exposure to residual surface contamination and soil contamination, respectively. It has been decided that use of Default Screening Values for residual levels of radiological contamination on structure surfaces and in soil will provide a conservative approach to assuring that the annual NRC dose criteria of 25 mrem from residual contamination in this facility is not exceeded. Appendix B presents the justification for use of Default Screening Values. Based on this justification, the conservative Default Screening Values (NUREG/CR-5512, Vol 3) for the potential radionuclide contaminants have been selected as acceptable final status criteria (i.e., DCGLs). An exception is Ag-108m, for which Default Screening Values are not provided. The DandD software was therefore used to develop Default Screening Values for Ag-108m, based on the relative dose factors of Ag-110m and Ag-108m, as presented in Federal Radiation Guides #11 and #12. The DCGL development process for Ag-108m is described in further detail in Appendix C. Default Screening values for surface soil are as follows:

Radionuclide Default Screening Value for Surface Soil (pCi/g)

Co-60 3.8 Ag-108m 8.2 Ag-110m 4.92 Cs-137 11.0 During the latter stages of remediation, three locations of potentially impacted subsurface soil were identified. These locations are:

1) Surrounding the foundation drain piping outside the east wall of the FNR building Basement.

2) Beneath the 1st Floor slab at the freight door between the FNR to the MMPP buildings.

3) Beneath a crack in the reactor support foundation on the Basement level.

The total volume of impacted soil at these three locations is estimated to be less than 50 M 3 . Potential contaminants included several radionuclides, which were not identified on surfaces or in the exposed surface soil. These were C-14 and Eu-152. Ag-110m, which was present on surfaces was no detected at significant concentrations in the subsurface soil. In anticipation of possible contamination in 10

January 2012 these locations, DCGL's for subsurface soil were developed for likely scenarios, using RESRAD. Appendix E describes the DCGL development process, including the exposure scenarios evaluated and modeling parameters. DCGL's for subsurface soil are listed below.

Radionuclide DCGL for Subsurface Soil (pCi/g)

Co-60 46 C-14 2500 Ag-108m 130 Eu-152 103 Cs-137 211 Satisfying surface contamination criteria will be demonstrated by measurement of gross beta activity. Appendix D describes development of a gross beta criterion for the FNR. The resulting gross-beta DCGL is 5125 dpm/100 cm 2.

Satisfying soil contamination criteria will be demonstrated by the sum-of-ratios approach. The sum of ratios (SOR) of gamma contaminant concentrations to their respective Default Screening Values or subsurface soil DCGL' s must therefore be < Unity (i.e., <1.0).

Section 4.6.4 provides details on the approaches to demonstrate that criteria have been met.

4.4 Quality Assurance Program A Final Status Survey Quality Assurance Project Plan (QAPjP), appropriate for implementing the final status survey and developing associated documentation, has been developed. That QAPjP incorporates the appropriate regulatory requirements applicable to the planning and conduct of radiological surveys necessary for the termination of the FNR license and the release of the site for unrestricted use.

4.5 Final Status Survey Approach The objective of the FSS is to demonstrate that remedial actions have been effective in removal/reduction of radiological materials and contamination, and that as-left radiological conditions satisfy the NRC-approved criteria for termination of the FNR License and for future use of the FNR facility without radiological restrictions. The FSS will be performed in accordance with 11

January 2012 guidelines and recommendations presented in NUREG-1757 and MARSSIM.

FSS activities will be performed by trained and qualified personnel, using properly calibrated equipment, sensitive to the potential contaminants, and following documented operating procedures. Appendix A of the QAPjP contains a list of the procedures, applicable to this FSS.

4.5.1 Classification by Contamination Potential For the purposes of guiding the degree and nature of final status survey coverage, areas are first classified as impacted, i.e., areas that may have residual radioactivity from licensed activities, or non-impacted, i.e., areas that are considered unlikely to have residual radioactivity from licensed activities. Non-impacted areas do not require further evaluation. For impacted areas MARSSIM identifies three classifications of areas, according to contamination potential.

" 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 radiation surveys) above the DCGL.

Examples include: site areas previously subjected to remedial actions; locations where leaks or spills are known to have occurred; and former waste storage areas.

" Class 2- Areas that have, or had prior to remediation, a potential for radioactive contamination or known contamination, but are not expected to exceed the DCGL. Examples include: locations where radioactive materials were present in unsealed form; potentially contaminated transport routes; areas handling low concentrations of radioactive materials; and 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, based on site operating history and previous radiation surveys. Examples include: buffer zones around Class 1 and Class 2 areas, and areas with a very low potential for residual contamination, but having insufficient information to justify a non-impacted classification.

Facility history (including the Historical Site Assessment and radiological monitoring conducted during characterization) and remedial activities are the bases for classification.

Following NRC approval of the Final Status Survey Plan, the UM may make changes to the classification of an area as long as the classification is changed to one of higher contamination potential. A license amendment pursuant to 10 CFR 50.90 shall be obtained if the change would decrease an area classification (i.e., impacted to non-impacted, Class 1 to Class 2, Class 2 to Class 3, or Class 1 to Class 3).

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January 2012 4.5.2 Identification of Survey Units Impacted areas are divided into survey units for implementing the FSS. A survey unit is a portion of a facility with common contaminants and contamination potential and contiguous surfaces or areas. Table 4-1 lists the survey unit areas suggested by MARSSIM for application at the FNR facility.

The area of individual survey unit will follow these suggested maximum sizes.

Per guidance in MARSSIM Section 4.6, special consideration will be provided for survey units with structure survey areas

• 10 m 2 and land survey areas *100 M 2 .

Because the number of data points for statistical tests (refer to Section 4.5.4 of this Plan) is unnecessarily large and not appropriate for smaller survey unit areas units, a minimum of 4 measurements (or samples) will be obtained from such areas, based on judgment, and compared individually with the DCGL's.

TABLE 4-1. MARSSIM - RECOMMENDED SURVEY UNIT AREAS Class Recommended Survey Unit Area Structures Land 2

1 up to 100 m 2 up to 2000 m 2

2 100 to 1000 m 2 2000 to 10,000 m 3 no limit no limit 2_

m . square meter Based on a historical assessment, preliminary survey data obtained in November 2002, the characterization survey in April 2003, and radiological monitoring conducted during remedial activities, a listing of facility areas that are currently expected to be included in the final status survey, is presented in Table 4-2, along with the estimated areas, anticipated contamination potential classifications, and the projected number of survey units within each area. This list of survey units differs slightly from the initial list, provided in the Decommissioning Plan; differences are primarily due to deletion of some survey units (e.g., pool walls, pits and sumps, and ventilation equipment), which were removed during remediation. Classifications and survey unit boundaries may change, based on results as the final status survey progresses. In accordance with Appendix 2 of NUREG-1700, changes in classification, resulting in a decrease in survey rigor, require NRC concurrence. Changes in survey unit classification or boundaries, will require that the survey of the affected area be redesigned and that the survey and data evaluation be repeated.

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January 2012 TABLE 4-2. FNR SURVEY AREAS AND FINAL STATUS SURVEY CLASSIFICATIONS Approx No. of Room or Description Class Surface Survey Comments Reference Area Area (Mi) Units FNR Basement Basement Floor North 1 100 1 EW Ceiling Beam HSA - Sec 6.1 and Serves as Divisor Sec 5.2.1.1 Basement Floor South 1 60 1 EW Ceiling Beam HSA - Sec 6.1 and Serves as Divisor Sec 5.2.1.1 (Excludes Sumps & Pits)

Basement Hold-up Tank Sump, 1 60 1 HSA - Sec 6.1, Sec Floor Sumps & Pits 5.2.1.1, and Characterization Pkg SP-020 Basement North End - Walls 1 89 1 HSA - Sec 6.1, Sec (N, E & W)and 5.2.1.1; Sec 6.2, and Columns Sec 5.2.2.1 Basement South End - Walls 1 60 1 HSA - Sec 6.2, Sec (S, E & W) and 5.2.2.1, and Survey Columns 2007-2439 Basement Ceilings 2 236 1 EW Ceiling Beam HSA - Sec 6.1, Sec (North & South) Divides North & South 5.2.1.1, Surveys Ceilings 2007-2356, 2007-2439, and 2007-2462 FNR First Floor 1101 Floor - SW Quadrant 1 102 1 HSA - Sec 6.1 and Sec 5.2.1.2 1101 Floor - NW 1 102 1 HSA - Sec 6.1 and Quadrant Sec 5.2.1.2 1101 Floor - NE Quadrant 1 108 1 HSA - Sec 6.1 and Sec 5.2.1.2 1101 Floor - SE Quadrant 1 84 1 Hatch Cover Included HSA - Sec 6.1 and Sec 5.2.1.2 1101 SW Quadrant West 1 95 1 Wall Remediation HSA - Sec 6.1 and of Pool - Upper & (Includes SW Freight Sec 5.2.1.2 Lower Walls + Door & Some of Storage Column 3 Port Wall) 14

January 2012 Approx No. of Room or Description Class Area Survey Comments Reference Area Ae (M2) Units 1101 NW Quadrant West 1 88 1 Wall Remediation HSA - Sec 6.1, Sec of North Stairwell - (Includes Some of 5.2.1.2, Survey 2007-Upper & Lower Storage Port Wall) 2379, and Survey Walls 2007-2388 1101 North Stairwell 1 82 1 HSA- Sec 6.1 and Enclosure - External Sec 5.2.1.2 Upper & Lower Walls + Pipe Chase 1101 NE Quadrant - 1 79 1 HSA - 6.1 and Sec Upper & Lower 5.2.1.2 Walls 1101 SE Quadrant - 1 89 1 HSA - Sec 6.1 and Upper & Lower Sec 5.2.1.2 Walls + Column 4 1101 South Wall East of 1 57 1 Includes Blue Freight HSA - Sec 6.1 and Pool / West of Door Between FNR & Sec 5.2.1.2 PML/FNR Metal PML Hot Cell Area -

Door - UpperA& Remediated Lower Walls 1103 Janitor's Closet 2 31 1 Survey 2007-2385 1101 South Pool Wall 1 52 1 HSA - Sec 6.1 and Sec 5.2.1.2 1101 Reactor Pool 1 59 1 HSA - Sec 6.1 Footprint 1101 Upper Pool Support 2 130 1 HSA - Sec 6.2 and Beams Facing Pool Sec 5.2.2.2 (3 beams) 1101 Ceiling & Beams Not 2 620 1 HSA - Sec 6.2 and Facing Pool Sec 5.2.2.2 1101 North Pool Support 1 24 1 Remediated Surveys 2007-2272 Beam Bottom and 2007-2450 Fagade 1101 E &W Pool Floor 1 15 1 Survey 2007-2273 Hatch Covers (Bottom Sides) 1101 Outside BP Floor 2 13 1 No Remediation Surveys 2007-2335 Hatch Cover Bottom Required and 2007-2336 FNR Second Floor 15

January 2012 Approx No. of Area RoorSurface SuvyCmet Description Class Area Survey Comments Reference (M )

2 Units 2101 Hallway / Corridor 2 189 1 HSA - Sec 6.2, Sec (All Surfaces) 5.2.2.3, and Survey 2007-2191 2111 Equipment Room - 1 87 1 HSA - Sec 6.1, Sec Floor 5.2.1.3, and Survey 2007-2456 2111 Equipment Room - 1 90 1 Wall Remediation HSA - Sec 6.1, Sec Lower Walls 5.2.1.3, and Survey 2007-2456 2111 Equipment Room - 2 225 1 HSA - Sec 6.2, Sec Upper Walls & 5.2.2.3, and Survey Ceilings 2007-2456 2109 All Surfaces 2 70 1 HSA - Sec 6.2 and Sec 5.2.2.3, and Survey 2007-2248 2108 All Surfaces 2 110 1 HSA - Sec 6.2, Sec 5.2.2.3, and Survey 2007-2044 2107 Floor & Lower Walls 1 50 1 Room Used as Sample HSA - Sec 6.1 and Handling & Counting Sec 5.2.1.3 Room 2107 Upper Walls & 2 55 1 HSA - Sec 6.1 and Ceiling Sec 5.2.1.3 2106 All Surfaces 2 114 1 HSA - Sec 6.2, Sec 5.2.2.3, and Survey 2007-2347 2105 Men's Restroom 2 94 1 HSA - Sec 6.2, Sec (All Surfaces) 5.2.2.3, and Survey 2007-2263 2104 Pipe Chase 3 10 1 Survey 2007-2337 (All Surfaces) 2103 Janitor's Closet 1 32 1 HSA - Sec 6.1 and (All Surfaces) Sec 5.2.1.3 2102 Women's Restroom 2 84 1 HSA - Sec 6.2, Sec (All Surfaces) 5.2.2.3, and Surveys 2007-1958, and 2007-2307 FNR Third Floor 16

January 2012 Approx No. of Room Areaor Dsrpin Cas Surface Description Class Area Survey Comments Reference (M 2) Units 3101 Hallway/Corridor 2 145 1 HSA - Sec 6.2, Sec (All Surfaces) 5.2.2.4, and Surveys 2007-1961 and 2007-2155 3102 Electronics Lab 2 170 1 HSA - Sec 6.2, Sec (All Surfaces) 5.2.2.4, and Survey 2007-1948 3103 Floor & Lower Walls 1 60 1 Former Fume Hood for HSA - Sec 6.1, Sec Handling Rx Irradiation 5.2.1.4, and Survey Samples 2011-0043 3103 Upper Walls & 2 56 1 HSA - Sec 6.1, Sec Ceilings 5.2.1.4, and Surveys 2007-1946 & 2011-0043 3104 Floor & Lower Walls 1 72 1 Former Lab & RAM Area HSA - Sec 6.1, Sec 5.2.1.4, Surveys 2007-2016 & 2011-0043 3104 Upper Walls & 2 75 1 HSA - Sec 6.2, Sec Ceiling 5.2.2.4, Surveys 2007-2016, 2010-0113, and 2011-0043 3106J Floor & Lower Walls 1 15 1 Identified as Room 3166 HSA - Sec 6.1, Sec in HSA. Janitor Closet & 5.2.1.4, and Survey New-Fuel Vault 2007-2188 3106J Upper Walls & 2 21 1 Identified as Room 3166 HSA - Sec 6.2, Sec Ceiling in HSA. Janitor Closet & 5.2.2.4, and Survey New-Fuel Vault 2007-2188 3108 & 3109 Floor & Lower Walls 1 74 1 Control Room & HSA - Sec 6.1 and Records Storage Sec 5.2.1.4 3108 & 3109 Upper Walls & 2 80 1 Control Room & HSA - Sec 6.2 and Ceiling Records Storage Sec 5.2.2.4 3101A Floor 1 70 1 Identified in D Plan as HSA - Sec 6.1, Sec Room 3110 (Hatch 5.2.1.4, and Survey Covers Not Included) 2011-0043 3101A Pool Floor South 1 98 1 HSA - Sec 6.1, Sec Wall - Floor Up to 5.2.1.4, and Survey Crane Beam (5.5 m) 2007-2116 17

January 2012 Approx No. of Room or Description Class Surface Survey Comments Reference Area ________(mn Area2) Unt Units 3101A Lower E, W, and N 1 51 1 HSA - Sec 6.1, Sec Walls 5.2.1.4, and Survey 2007-2058 3101A Upper E, W, and N 2 423 1 HSA - Sec 6.2, Sec Walls, S Wall Above 5.2.2.4, and Survey Crane Support, and 2007-2116 Ceiling 3101A Pool Floor - East & 1 32 1 Attachment Points on HSA - Sec 6.1, Sec West Hatch Covers West Hatch Covers 5.2.1.4, and Survey (Tops) & Contaminated 2007-2034 Penetrations FNR Crane 2 60 1 Class 2 Based on Characterization Pkg 3101A Survey Results SP-014 and Survey 2007-2478 FNR Fourth Floor Fourth Floor Cooling Tower 3 630 1 Includes Splashboard HSA - Sec 6.2, Sec Structure Unit 5.2.2.5, and Characterization Pkg SP-016' Fourth Floor FNR Stack Plenum 1 100 1 Remediated HSA - Sec 6.1 and Sec 5.2.1.4 FNR Stairs 1 Stair No. 1 South Stairwell 3 292 1 HSA - Sec 6.2 and Survey 2007-2315 Stair No.1 South Stairwell to 2 35 1 HSA - Sec 6.1, Sec Basement 6.2, and Survey 2007-2315 Stair No. 2 North Stairwell 3 273 1 HSA - Sec 5.2.2.1, and Survey 2007-2329 Stair No. 3 Cooling Tower - 1 20 1 Former RAM storage. HSA - Sec 6.1 and Stairwell Entrance No Remediation Sec 5.2.1.4 and Remaining Required.

Stairs Stair No. 3 Cooling Tower - 2 180 1 HSA - Sec 6.1, Sec Upper & Lower 5.2.1.4, and Survey Walls 2007-2245 18

January 2012 Approx No. of RoomAor Description Class Surface Survey Comments Reference Area Area 2

(M ) Units FNR Roof &

Outside Areas FNR Exterior Walls and Roof 3 454 1 Includes Doors, Vents, HSA - Sec 5.2.2.5 and Stacks and Characterization Pkg SP-017 Outdoors Temporary 1 150 2 Storage Pad Area -

Decommissioning West Side - Temporarily Lay Down Area / Fenced-off as D&D West Side (Soil, Shipment Staging Area Concrete, and BP Hatch Covers)

Pipe Chases Cooling Supply & Return 3 110 1 HSA - Sec 6.2, Sec Tower Vertical Pipe Chase 5.2.1.5, and Sec 5.2.2.5 2 nd Floor Pipe Chase Behind 3 60 1 Based on operational Room 2103 history and surveys Drains Storm Drains Storm Drains 1 3 N/A 1 Characterization Pkg SP-012 Sanitary Sanitary,Sewer 2 N/A 1 HSA - Sec 6.1, Sec Sewer 5.3, and Characterization Pkg SP-012 Drain Tile Foundation Drain 1 35 1 Decom Plan Sections HSA - Sec 5.3 and Pipe Tile (East Side) 2 2.1.1 and 2.1.2.4 Sec 6.1 Miscellaneous Room 1101 Storage Ports 1 13 1 Underground / Characterization Pkg (Truncated / Internal Subsurface Storage SP-004 Surfaces) Ports Removed from West Side Soil Subsurface Below FNR 1 12 m3 Exposed Soil After Characterization Pkg Soil Basement & BP (-73 M) Removing Embedded SP-018 Floor Foundation Piping Subsurface Foundation Drain 1 35 m3 Soil Sampling & HSA - Sec 6.1 and Soil Tile (East Side) 3 (-54 M2) Analyses Completed Sec 5.3 19

January 2012 Approx No. of Room or Description Class Surface Comments Reference Area Area Survey 2

(M ) Units Subsurface Excavated/Removed 1 1.5 m3 Activated Soil HSA - Sec 5.2.2.2 Soil Storage Ports Remediated After (West Side) 4 Removal of Storage Ports (Survey Complete)

Footnotes:

(1) Addressed during Site Characterization (Characterization Package SP-012) in December 2007.

Results will be provided in final report.

(2) Final decision on the disposition of the foundation drain tile pipe will be provided in the final report.

(3) Soil sampling in the vicinity of the foundation drain tile pipe (east side) was conducted in September 2010 & September 2011. Results will be provided in the final report.

(4) Soil sampling in the vicinity of the FNR storage ports (west side) was completed in December 2008

/ January 2009. Results will be provided in the final report.

Note: BP - Beam Port Floor (Room 1101)

HSA - Historical Site Assessment (CH2MHill / January 2003) 4.5.3 Testing to Demonstrate Compliance The Null Hypothesis for the statistical test to demonstrate compliance with project criteria is "Residual FNR radiological contamination levels exceed project criteria". The objective of the FSS is to reject this Null Hypotheses, by demonstrating at a Type I (a) decision error level of 0.05 (i.e., 95% confidence level) that the contamination does not exceed criteria. The Type II (P) decision error level is also 0.05. Because there are multiple potential contaminants in soil, compliance with soil criteria will be evaluated using the sum-of-ratios (SOR) approach. There are multiple building surface types (concrete, metal, wood, glass, etc.) in most survey units and individual survey unit measurements will be adjusted for appropriate material background contributions, using the paired-data approach. For both the soil and building surface surveys, the sign test is the appropriate statistical test of compliance.

4.5.4 Survey Data Requirements To establish the number of data points needed to demonstrate that residual contamination criteria have been satisfied, a parameter known as the "relative shift", which effectively describes the distribution of final sample data, is calculated, as follows:

(1) A/u = (DCGL-LBGR)/ o 20

January 2012 where:

A/o = relative shift DCGL = cleanup level (Section 4.3).

LBGR = lower bound of the gray region and is defined in the DQOs as 50 percent of the DCGL. Where final sample data are not yet available, MARSSIM guidance (Section 5.5.2.2) assigns a value of one-half of the DCGL for the LBGR.

o = standard deviation of the sample concentrations in the survey unit. Where final sample data are not yet available, MARSSIM guidance (Section 5.5.2.2) is to use a value of 30 percent of the DCGL.

Using the equation for relative shift and MARSSIM guidance for situations where final sample data are not yet available, the relative shift for design purposes is (1 - 0.5)/0.3 for a value of 1.67. Based on the relative shift of 1.67 and Type I and Type II decision errors of 0.05, the number of required data points from each survey unit to perform the sign test, as obtained from MARSSIM guidance (Table 5.5) is 17.

Once actual sample data are collected, the MARSSIM DQO process requires a retrospective assessment of the selected LBGR and a values, to confirm that an adequate number of data points was obtained for final evaluation.

4.5.5 Survey Locations MARSSIM recommends a random-start systematic triangular measurement or sampling pattern for FSS of Class I and Class 2 survey units. This type of triangular pattern will be used for this final status survey, except where dimensions and/or other factors related to a specific survey unit require use of an alternate pattern. The spacing (L) between data points on a triangular pattern is determined by:

L = [(Survey Unit Area)/ (0.866 x number of data points)11/ 2 To simplify the designation of data points while assuring a sufficient number of data points are obtained for statistical purposes, the value of L is rounded to the nearest whole meter. If the systematic pattern does not provide sufficient data points to satisfy the number determined in Section 4.5.4, additional data points will be identified, using a random-number technique.

For FSS of Class 3 survey units, measurement or sampling data points will be judgmental, based on professional opinion. These data points will be biased to locations considered to have the highest probability of residual contamination, with additional locations chosen to provide distributed survey unit coverage.

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January 2012 4.5.6 Survey Design Packages FSS designs will be prepared and documented in a survey design package.

Multiple survey units, having similar history, classification, and conditions, may be covered by one design package. These design packages will include survey unit maps and drawings, classification, scan frequency, data point locations, a description of unusual/unique conditions that may require deviation from standard survey techniques, and alternative techniques to be used in such cases.

4.5.7 Survey Instrumentation Table 4-3 is a list of radiological survey instrumentation that will be used to implement the FNR FSS.

These instruments are maintained and calibrated in accordance with UM procedures HP-211 and HP-402. For simplicity in application to FSS, instrument response (efficiency) is based on NIST-traceable sources of Tc-99 (beta EMAX =

292 keV) and Th-230 (alpha E = 4.68 MeV). The energies of these radionuclides are representative of the dominant potential contaminants and thus will provide conservative overestimates of the contaminant mixture. For field measurement applications, calibration represents 2ri response. Effects of surface conditions on measurements are integrated into the overall instrument response through use of a "source efficiency" factor, in accordance with the guidance in ISO-7503-1, Evaluation of Surface Contamination - Part 1: Beta Emitters and Alpha Emitters (First Edition) and NUREG/CR-1507, Minimum Detectable Concentrationswith Typical Radiation Survey Instruments for Various Contaminantsand Fields Conditions.

Default source efficiency factors, of 0.5 for beta-emitters > 0.4 MeV Emax and 0.25 for beta-emitters between 0.150 MeV and 0.400 MeV Emax (per ISO-7503-1) are generally applicable to anticipated FNR contaminants and surface conditions.

For the predominant maximum beta energy of approximately 0.300 MeV from Co-60, a source efficiency value of 0.37 will be used. If contaminants or conditions in specific survey units are not consistent with use of these default values, specific instrument response and source efficiency factors will be determined and documented in the final status survey packages for those survey units.

Detection sensitivities are estimated using the guidance in MARSSIM and NUREG-1507, Minimum Detectable Concentrationswith Typical Radiation Survey Instrumentsfor Various Contaminantsand Fields Conditions (NRC, 1997b).

Instrumentation and survey techniques are chosen with the objective of achieving detection sensitivities of < 50% of the criteria for structure surfaces, for both scanning and direct measurement. This assures identification of areas potentially exceeding the established project criteria.

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January 2012 TABLE 4-3. INSTRUMENTATION FOR FNR FINAL STATUS SURVEY Detector Type Make Meter Application Sensitivity (dpm/100 cm 2 , except as noted)

Scanning Static Count (1 minute)44-142 Beta Scintillation Ludlum 2221 Beta scan and 1770 560 measurement DCGL = 5125 DCGL = 5125 43-37 Gas Proportional Ludlum 2221 Beta scan 460 N/A Floor/Wall DCGL = 5125 Monitor 44-10 Nal Ludlum 2221 Gamma scan 2.62 pCi/g Co-60; N/A DCGL = 3.8 pCi/g 1.39 pCi/g Ag-108m; DCGL = 8.2 pCi/g 5.08 pCi/g Cs-1 37; DCGL =11.0 pCi/g 1.49 pCi/g Ag-1 1Om; DCGL = 4.92 pCi/g (surface soil)

Tennelec LB5100 Gas proportional Tennelec N/A Alpha smear N/A 38 measurement DCGL N/A (no a contaminants)

Tennelec LB5100 Gas proportional Tennelec N/A Beta smear N/A 38 measurement DCGL = 512 cm2 - square centimeter, dpm - disintegrations per minute, g - gram., pCi - picocuries.

Measuring instruments are calibrated at prescribed time periods or usage and whenever the accuracy of the equipment is suspect. Calibration is performed using standards traceable to NIST or an equivalent standard organization.

Instruments are suitably marked or otherwise identified to indicate calibration status. Instruments found to be out of calibration, require a documented evaluation, commensurate with the significance of the condition, of the validity of data obtained with that instrument since its previous acceptable performance.

Instruments are properly handled and stored to maintain accuracy and shall follow ANSI N323B-2003.

Operational and background checks will be performed at the beginning and end of each day of final status survey activity and whenever there is reason to question instrument performance. These checks should follow procedure HP-401.

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January 2012 4.5.8 Background and Reference Area Measurements In addition to the instrumentation background response, many construction materials and environmental media (e.g., soil, sediment) contain naturally occurring levels of radioactive materials, which contribute to a survey measurement. Background contributions must therefore be determined, if 1) the residual contamination includes a radionuclide that occurs in background, or 2) measurements are not radionuclide-specific. Multiple background determinations will be required for the final status survey. A set of reference measurements must be obtained for each instrument being used for survey unit evaluation. For applications involving the Sign test, sufficient background determinations will be made for each media or surface material and with each instrument to provide an average background level that is accurate to within +/-

20 percent; this usually requires 8 to 10 measurements, which are then evaluated using the procedure described in draft NUREG/CR-5849, Manualfor Conducting Radiological Surveys in Support of License Termination (NRC, 1992a), and additional data points obtained, as necessary. Reference area and background requirements will be identified at the time of individual survey unit final status survey design.

With several exceptions, structural material backgrounds will be obtained from surfaces in the Cooley Building Tunnel, which is of similar construction and age as the FNR Building, but without a history of radioactive materials use. Because there are no other known sources of high-density concrete in University facilities, background measurements for this material will be obtained at ceiling and upper wall locations in areas of the FNR, where use history indicates a negligible potential for impact by facility operations. Another structural material for which a source outside the FNR has not been identified is the glazed tile in janitor closets, restrooms, and change rooms. It is proposed to obtain background measurements for this material at ceiling and upper wall locations in the 2nd floor women's restroom, where the potential for impact by facility operations is considered negligible. Bulk samples of high-density and glazed tile material from these locations will be analyzed to confirm the absence of other than naturally occurring radioactive material.

4.5.9 Survey Reference Systems A grid system will be established on surfaces to provide a means for referencing measurement and sampling locations. On Class 1 and 2 structure surfaces, a 1-m interval grid will be established; a 5-m interval grid will be established on Class 3 structure surfaces; and a 10-m interval grid will be established for land area surfaces. Grid systems typically originate at the southwest corner of the survey unit, but specific survey unit characteristics may necessitate alternate grid origins. Grids are assigned alphanumeric indicators to enable survey location identification. Structure grids are referenced to building features; open land

.grids are referenced to the state or federal planar grid system. Maps and plot 24

January 2012 plans of survey areas will include the grid system identifications. Systems and surfaces of less than 20 m 2 will not be gridded, but survey locations will be referenced to prominent facility features. Procedure FSS-02 describes mapping and gridding for FSS purposes.

4.5.10 Survey Techniques Data collected for final status survey of structure surfaces will consist of scans to identify locations of residual contamination, direct measurements of beta surface activity, and measurements of removable beta surface activity. Final status survey of open land (soil) areas will consist of scans to identify locations of residual contamination and samples of soil, analyzed for potential contaminants.

Additional measurements and samples will be obtained, as necessary, to supplement the information from these typical survey activities. Survey techniques are described in more detail in this section.

4.5.10.1 Beta Surface Scans Beta scanning of structure surfaces will be performed to identify locations of residual surface activity. Gas-flow proportional detectors and scintillation detectors will be used for beta scans. Floor monitors with 580 cm 2 gas proportional detectors will be used for floor and other larger accessible horizontal surfaces; hand-held 100 cm 2 scintillation detectors will be used for surfaces not accessible by the floor monitor. Scanning will be performed with the detector within 0.5 cm of the surface (if surface conditions prevent this distance, the detection sensitivity for an alternate distance will be determined and the scanning technique adjusted accordingly). Scanning speed will be no greater than 1 detector width per second. Audible signals will be monitored and locations of elevated direct levels identified for further investigation.

Minimum scan coverage will be 100 percent for Class 1 surfaces, 25 percent for Class 2 surfaces, and 10 percent for Class 3 surfaces. Coverage for Class 2 and Class 3 surfaces will be biased towards areas considered by professional judgment to have highest potential for contamination.

4.5.10.2 Gamma Surface Scans Gamma scanning surfaces will be performed on structure and soil surfaces to identify locations of residual surface activity. Nal gamma scintillation detectors (2 inch x 2 inch) will be used for these scans. Scanning will be performed by moving the detector in a serpentine pattern, while advancing at a rate of approximately 0.5 m per second. The distance between the detector and the surface will be maintained within 5 cm of the surface. Audible signals will be monitored and locations of elevated direct levels identified for further investigation.

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January 2012 Minimum scan coverage will be 100 percent for Class I surfaces, 25 percent for Class 2 surfaces, and 10 percent for Class 3 surfaces. Coverage for Class 2 and Class 3 surfaces will be biased towards areas considered by professional judgment to have highest potential for contamination.

4.5.10.3 Surface Activity Measurements Direct measurement of beta surface activity will be performed at designated locations using a 100-cm 2 plastic scintillation detector. Measurements will be conducted by integrating the count over a 1-minute period. Where adverse surface conditions may result in underestimating activity by direct measurements, surface samples will be obtained for laboratory analyses. Need for such sampling will be identified in final status survey design for specific survey units.

4.5.10.4 Removable Activity Measurements A smear for removable activity will be performed at each direct surface activity measurement location. A 100 cm 2 surface area will be wiped with a 2 inch diameter paper filter or cloth, using moderate pressure. Smears will be analyzed onsite for gross alpha and gross beta activity using a Tennelec gas proportional automatic sample counter.

4.5.10.5 Soil Sampling Samples of surface (upper 15 cm) soil will be obtained from selected locations using a hand trowel or bucket auger. Approximately 500 to 1000 g of soil will be collected at each sampling location.

4.5.10.6 Special Situations There will likely be several areas that do not meet the definition of exposed soil or structure surfaces. Examples include the foundation drain around the exterior foundation and the soils in the immediate vicinity of the removed source storage tubes. Where such special situations are encountered, survey approaches and evaluation methods will be developed on a case by case basis and described in the survey design package.

4.6 Data Evaluation and Interpretation 4.6.1 Sample Analysis Smears for removable activity will be analyzed by the onsite laboratory for gross alpha and gross beta activity. Soil will be screened onsite for gamma emitters and, if screening indicates the sum-of-ratios for the four contaminants of concern (see Section 4.3) is less than unity, samples will be analyzed at the commercial 26

January 2012 offsite laboratory by gamma spectrometry for final evaluation that decommissioning criteria have been satisfied.

4.6.2 Data Conversion Measurement data will be converted to units of dpm/100 cm 2 or pCi/g for comparison with guidelines and/or for statistical testing. Where appropriate for Sign tests, data will be adjusted for material and instrument background contributions.

4.6.3 Data Assessment Data will be reviewed to assure that the type, quantity, and quality are consistent with the survey plan and design assumptions. Data standard deviations will be compared with the assumptions made in establishing the number of data points.

Individual and average data values will be compared with guideline values and proper survey area classifications will be confirmed. Individual measurement data in excess of the guideline level for Class 2 areas and in excess of 25 percent of the guideline for Class 3 areas will prompt investigation. Patterns, anomalies, and deviations from design assumption and plan requirements will be identified.

Need for investigation, reclassification, remediation, and/or resurvey will be determined; a resolution will be initiated and the data conversion and assessment process repeated for new data sets.

4.6.4 Determining Compliance with Guidelines 4.6.4.1 Sign Test For a structure surface survey unit to be evaluated using the Sign test, individual activity values and the average and standard deviation activity values will be calculated.

If all individual values for a survey unit are less than the guideline level, that survey unit satisfies the criterion and no further evaluation is necessary; the null hypothesis is rejected, and the survey unit meets the established criteria.

If any individual value is greater than the guideline value, the null hypothesis is accepted, and the survey unit does not meet the established criteria; investigation, remediation, reclassification, and/or resurvey will be performed, as appropriate.

4.6.4.2 Unity Rule Sign Test For an open land or structure surface survey unit to be evaluated using the Unity Rule Sign test, individual activity values and the ratios of the activity values to their respective guideline values will be calculated. For each data location add the ratios together to determine the Sum of Ratios.

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January 2012 If all Sum of Ratios values for the survey unit are less than 1, that survey unit satisfies the criterion and no further evaluation is necessary; the null hypothesis is rejected, and the survey unit meets the established criteria.

If the Sum of Ratios value for any sample is greater than 1, the survey unit does not satisfy the criterion. The null hypothesis is accepted, and investigation, remediation, reclassification, and/or resurvey will be performed, as appropriate.

4.7 Isolation and Control Measures Following completion of FSS, the survey unit will be isolated and access controlled. Routine access, equipment removal, material storage, and worker and material transit through the area without proper controls are no longer allowed. One or more of the following administrative and physical controls will be established to minimize the possibility of introducing radioactive material from ongoing decommissioning activities in adjacent or nearby areas:

" Personnel training,

• Installation of barriers to control access to the area(s),

• Installation of postings with access and egress requirements, and/or

" Locking or otherwise securing.

Isolation and control will be discontinued following NRC acceptance that the project decommissioning criteria have been satisfied.

4.8 Final Status Survey Report A report describing the survey procedures and findings will be prepared for submission to the NRC in support of license termination. The survey report will provide a complete record of the facility's radiological status and a comparison to the site release criteria. The survey report will include survey data and overall conclusions, which demonstrate that the FNR Facility meets the radiological criteria for unrestricted use. Information such as the number and type of measurements, basic statistical quantities, and statistical test results will be included in the report. The survey report will contain additional detail to enable an independent or third party re-creation and evaluation of the survey results and a determination as to whether the site release criteria have been met.

The following outline illustrates a general format that may be used for the final status survey report and may be adjusted to provide a clearer presentation of the information. The level of detail will be sufficient to clearly describe the final status survey program and certify the results.

Information to be submitted:

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January 2012

• A summary of the results of the final status survey,

• A discussion of any changes that were made in the final status survey from what was proposed in the LTP or other prior submittals,

• A description of the method by which the number of samples were determined for each survey unit,

• A summary of the values used to determine the numbers of samples and a justification for these values (refer to Section 4.5.4),

" The results for each survey unit including:

1. Number of samples taken for the survey unit.

2. A map or drawing of the survey unit showing the reference system and random start systematic sample locations for Class 1 and 2 survey units, and random locations shown for Class 3 survey units and reference areas.

3. Measured sample concentrations.

4. Statistical evaluation of the measured concentrations.

5. Judgmental and miscellaneous sample data sets reported separately from those samples collected for performing the statistical evaluation.

6. Discussion of anomalous data including any areas of elevated direct radiation detected during scanning that exceeded the investigation level or measurement locations in excess of the DCGLw.

7. A description of follow-up actions and results.

8. A statement that a given survey unit satisfied the DCGLw.

• A description of any deviations from initial survey design and survey techniques, and

" A description of the investigation and follow-up actions when the FSS fails to demonstrate that the criteria have been satisfied.

4.9 References Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), NUREG-1575 (Rev. 1), US Nuclear Regulatory Commission, 2000.

University of Michigan Decommissioning Plan for the Ford Nuclear Reactor (Rev 1.),

University of Michigan, January 2006.

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January 2012 Consolidated NMSS Decommissioning Guidance. NUREG-1757, US Nuclear Regulatory Commission, 2000.

Residual Radioactive Contamination from Decommissioning, NUREG/CR-5512 Vol. 3, US Nuclear Regulatory Commission, 2000.

Manual for Conducting Radiological Surveys in Support of License Termination, NUREG/CR-5849 (draft), US Nuclear Regulatory Commission, 1992.

Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, NUREG/CR-1507, US Nuclear Regulatory Commission, 1997.

Evaluation of Surface Contamination - Part 1: Beta Emitters and Alpha Emitters, ISO-7503-1, International Organization for Standardization, 1988.

Installed Radiation Protection Instrumentation Test and Calibration - Portable survey Instruments for Near Background Operation, ANSI-N323B-2003, American national Standards Institute, 2003.

Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, Federal Guidance Report No. 11, EPA-520/1-88-1988, Environmental Protection Agency, 1988.

External Exposure to Radionuclides in Air, Water, and Soil, Federal Guidance Report No. 12, EPA-402-R-93-081, Environmental Protection Agency, 1993.

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January 2012 Appendix A Identification of Residual Radiological Contaminants at the Remediated Ford Nuclear Reactor Facility Numerous radionuclides were potential contaminants at the FNR as a direct result of reactor operations (i.e., fission and activation products) as well as experiments performed in the facility. Research of historical documents, interviews, and preliminary characterization of the facility identified the radionuclides potentially present at the time decommissioning was initiated as Sb-125, C-14, Cs-137, Co-60, Eu-152, Eu-154, Fe-55, H-3, Mn-54, Ni-63, Ag-108m, Ag-110m, and Zn-65. Monitoring of waste and the remaining structure during the remedial activities identified Ba-133 (from activation of high-density barytes concrete) as an additional potential contaminant.

As part of the decontamination and dismantlement efforts, items and structural components in close proximity to the reactor core, including most of the high-density barytes concrete bioshield, were removed. Direct monitoring and sampling of surfaces after remediation was complete, indicated little, if any, residual surface contamination.

Table A-1 presents a summary of analyses of samples, representative of as-left surface conditions.

Table A-1 Analyses of Post-Remediation Samples Radionuclide Samp Samp Samp Samp Samp Samp Samp 1 2 3 4 5 6 7 (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g)

H-3 <2.04 <2.17 <2.17 <2.21 <2.11 0.62 1.49 C-14 <1.09 <1.14 <1.08 <1.12 <1.07 <0.47 1.26 Fe-55 <7.72 <8.17 <11.7 <10.4 <9.03 <20.1 <17.4 Co-60 <0.05 0.55 0.43 4.21 <0.08 <0.03 <0.03 Ag-108m <0.05 0.30 0.14 0.36 <0.08 <0.03 <0.03 Ag-110m <0.07 <0.12 <0.19 <0.27 <0.06 <0.07 <0.05 Cs-137 <0.04 <0.07 0.24 1.77 <0.07 0.02 <0.03 Ba-133 <0.06 0.11 <0.13 <0.14 <0.09 <0.10 <0.24 Mn-54 <0.07 <0.12 <0.19 <0.24 <0.08 <0.07 <0.16 Eu-152 <0.32 <0.55 <0.91 <0.72 <0.54 <0.59 <1.03 Sample 1: UM-2009-06-08-01; Concrete grindings from foundation under North side of former reactor pool Sample 2: UM-2009-06-10-01; Concrete grindings from foundation under West side of former reactor pool Sample 3: UM-2009-06-12-01; Concrete grindings from foundation under former reactor thermal column Sample 4 UM-2009-06-19-01; Concrete grindings from foundation directly beneath former reactor core 31

January 2012 Sample 5: UM-2010-01-29-01; Concrete grindings from foundation under West side of former reactor pool Sample 6: UM-2010-05-18-01; Concrete grindings from foundation under West side of former reactor pool; after additional remediation of area identified by gamma scan Sample 7: UM-2010-05-27-01; Concrete grindings from foundation under West side of former reactor pool; after additional remediation of area identified by gamma scan It is evident from these analyses of post-remediation samples that only a few samples contained positive concentrations of radionuclides, attributable to FNR operations.

Most concentrations were either less than laboratory detection limits or very low, relative to typical DCGL levels. Co-60, Ag-108m, and Cs-137 are the only radionuclides identified in more than 2 of the 7 samples and/or at concentrations that could be considered positive indication of their presence. Because of the high fraction of non-detectable levels, these data cannot be used to develop meaningful radionuclide ratios for the remediated facility.

Several samples were obtained during remediation from locations contaminated by leakage of reactor pool water and thus believed representative of potential surface contamination throughout the FNR facility. Table A-2 presents a summary of analytical results for these samples. It should be noted that, based on the absence of non-gamma-emitting radionuclides in the post-remediation samples, these analyses included only gamma spectrometry.

Table A-2 Analyses of Residue Samples Radionuclide* Sample Sample Sample Sample Average 8 9 10 11 Fraction (pC/g) (pCi/g) (pCig) (PC§/g)

Co-60 37.7 12.1 0.35 4.6 0.312 Ag-108m 71.5 29.3 2.3 <0.03 0.588 Ag-110m 5.87 2.09 <0.06 <0.46 0.048 Cs-137 1.89 1.50 0.20 <0.03 0.021 Mn-54 0.28 <0.14 <0.02 <0.02 0.003 Zn-65 1.56 0.48 0.05 <0.05 0.012 Ba-133 1.07 0.92 <0.04 <0.04 0.012 Eu-152 <0.29 <0.30 <0.06 <0.06 0.004

• Analyses were by gamma spectrometry and did not include hard-to-detect radionuclides. No other gamma-emitting contaminants were detected.

Sample 8: UM-2009-08-10-01; Concrete grindings from core A in floor slab crack.

Sample 9: UM-2009-10-01-01; Concrete grindings from core L in floor slab crack.

Sample 10: UM-2010-09-16-D823-821; Material from Drain Tile.

Sample 11: UM-2010-08-27-02; Soil beneath MMPP/FNR freight door.

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January 2012 The fractional contributions of the radionuclides in samples 8-11 were calculated and the average contribution for these four samples were determined. Results are presented in Table A-2. To maximize potential contributions to future occupant dose for surrogate determination, where the activity was reported as less than the minimum detectable activity (MDA), the MDA value was used in calculating the activity ratios. Appendix D describes application of these ratios in establishing a gross beta DCGL for surface contamination at FNR.

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January 2012 Appendix B Justification for Use of Default Screening Values for FNR Soils and Surfaces In accordance with Section 6.6.6 of NUREG-1757, the following conditions must be satisfied to justify use of the Default Screening Values:

Building Surface Contamination I. The contamination on building surfaces should be surficial and non-volumetric (e.g., <10 mm (0.4 in).

Justification: Paint and tile have been removed to expose potentially contaminated structure surfaces. Fixtures and portions of the structural concrete in proximity to the reactor core, where activation may have resulted in volumetric contamination, have been removed. Samples of remaining concrete, representing material at 0.5 to 2.0 and 2.0 to 4.0 inches below the surface, were obtained at two locations, which are closest to the former position of the reactor core. Analyses of these samples did not identify any detectable radionuclides of reactor origin (see Table B-1), indicating absence of volumetric contamination.

2. Contamination on surfaces is mostly fixed (not loose), with the fraction of loose contamination not to exceed 10 percent of the total surface activity.

Justification: Routine surveys, during and following remediation, did not identify the presence of removable contamination. Smears for removable contamination will be obtained at direct measurement locations during the FSS. Removable contamination must be less than 10% of the gross beta DCGL.

3. The screening criteria may not be applied to surfaces as buried structures (e.g.,

drainage or sewer pipes) or mobile equipment within the building; such surfaces and buried surfaces will be treated on a case-by-case basis.

Justification: Potentially impacted piping, including imbedded piping and adjacent concrete, has been removed, with exception of the short section of piping to the French Drain on the Basement level. Survey and evaluation of the drain system will be performed separately.

• Surface Soil Contamination

1. The initial residual radioactivity (after decommissioning) is contained in the top layer of the surface soil (e.g., approximately 15 cm (6 in)).

Justification: A small volume of soil around a storage tube in the west wall of the first floor level, used to store several PuBe neutron sources, was removed and scans and sampling of the remaining soil surfaces was performed. That evaluation has been completed and results will be provided separate from the 34

January 2012 remaining facility FSS. The foundation drain, around the foundation of the FNR received a short-term release of low-level contaminated water.

Sampling of soil from that system has been performed and results will be provided separate from the remaining facility FSS. Otherwise, potential contamination of soil is limited to soil surfaces on the basement level, exposed during removal of impacted imbedded piping, and soil beneath the first floor at the doorway (South West Freight) between the FNR and MMPP Buildings, resulting from seepage of pool water through a small crack in the bioshield and a possible liquid spill on the floor above the gap between the FNR/MMPP buildings. Consequently, the source of any contamination of soil, other than that in the vicinity of the storage tubes, foundation drain system, and gap between the FNR and MMPP will be due to dispersal of contamination from structure surfaces and will be limited to the exposed soil surfaces.

2. The unsaturated zone and the ground water are initially free of contamination.

Justification: As part of the initial facility characterization in 2007, a ground water monitoring well, installed near the entrance to the MMPP, which is downstream from the FNR facility, was sampled. Sampling of that well (Table B-2) did not identify evidence that the ground water has been impacted as a result of FNR operations.

3. The vertical saturated hydraulic conductivity at the specific site is greater than the infiltration rate.

Justification: No residual radiological contamination of subsurface soil, due to past FNR operations, has been identified. The elevations of the soil in the vicinity of the source storage tubes, cavity beneath the MMPP/FNR SW freight door, and French drain are approximately 839, 829, and 823 ft.,

respectively. Elevation of the upper bound of the saturated zone is 809 ft.

The vertical distance from contaminated soil to the water table is therefore at least 14 ft. There are currently no liquid discharges from the FNR to the ground water in the vicinity of this facility.

Based on results of monitoring in support of remedial actions and ongoing radiological surveys during and following remediation, the above criteria are satisfied. Therefore, the conservative Default Screening Values (NUREG/CR-5512, Vol 3) for the potential radionuclide contaminants have been selected as acceptable final status criteria (i.e.,

DCGLs).

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January 2012 Table B-1 Analyses of Concrete Cores Radionuclide Concentration (pCi/g) 2010-06-17-02 2010-06-17-03 2010-06-17-04 0.5 - 2.0 inches 0.5 - 2.0 inches 0.5- 2.0 inches Co-60 <0.01 <0.03 <0.03 Ag- 108m <0.01 <0.03 <0.03 Ag- 110m <0.01 <0.03 <0.03 Cs-137 <0.01 <0.04 <0.03 Mn-54 <0.08 <0.04 <0.03 Zn-65 <0.04 <0.11 <0.10 Ba-133 <0.03 <0.07 <0.07 Eu-152 0.03 <0.08 0.11 Table B-2 Analyses of Ground Water Samples from Well Near PML Entrance Radionuclide Concentration (pCi/l) 7/2/2007 8/8/2007 Gross alpha 58.2 + 9.7 20.1 + 6.5 Gross beta 67.2+8.4 31.3 +7.1 H-3 <706 <648 Co-60 <6.68 <9.88 Cs-137 <5.69 <8.07 Ag-108m <6.03 <7.31 Ag-110m <5.28 <8.53 Mn-54 <7.03 <8.85 Zn-65 <14.1 <20.6 Eu- 152 <42.0 <53.4 36

January 2012 Appendix C Development of Default Screening Values for Ag-108m Default Screening Values, listed in NUREG/CR-5512, Volume 3, do not include the radionuclide Ag-108m. Also, the DandD code, used to develop those Default Screening Values, does not include the dose values for that radionuclide. Therefore the following approach was used to develop Default Screening Values for Ag-108m:

1. DandD Version 2.2.0 was used to calculate contributions of individual pathways (i.e., by excluding all other pathways) to the dose (mrem/y) from 1000 dpm/100 cm 2 Ag-110m for the Building Occupancy scenario and from 1.0 pCi/g for the Residential scenario. The surface activity and soil concentrations of Ag-110m, consistent with a 25 mrem/y dose, were then calculated to confirm that the calculation was performed correctly. The resulting calculations yielded 10205 dpm/ 100 cm 2 and 4.87 pCi/g, as compared with the values of 1.02E+4 dpm/100 cm 2 and 4.92 pCi/g, respectively, in NUREG/CR-5512, Volume 3. (Refer to attached table)

2. These individual pathway doses for Ag-110m were adjusted by multiplying the Ag-110m contributions by the Ag-108m to Ag-110m ratios of the following dose values for external and internal exposure, as presented in FRG #12 and FRG #11, respectively.

Pathway Ag-110m Dose Ag-108m Dose External - surface 2.65E-15 Sv/Bg s im-2 1.60E-15 Sv/B s im-2 External - 15 cm thick 7.93E-17 Sv/Bg s im-3 4.61E-17 Sv/Bg s im-3 Inhalation 2.17E-08 Sv/Bg 1.07E-08 Sv/Bg Ingestion 2.92E-09 Sv/Bg 2.06E-09 Sv/Bg

3. The resulting individual Ag-108m doses for all pathways were then calculated and the surface activity and soil concentration values consistent with a 25 mrem/y dose were calculated, yielding Default Screening Values of 1.70E+04 dpm/100 cm 2 and 8.20 pCi/g for building surface and surface soil contamination, respectively.

37

January 2012 Scenario Pathway Ag-110m Dose Dose factors Ag-108m Dose (mrem/1000 dpm/100 cm 2) (Ag-108m/Ag-110m) (torem/1000 dpm/100 cm 2)

Building Occupancy External Exposure 2.34E1+00 0.604 1.41E+00 Inhalation 1.05E-01 0.493 5.18E-02 Ingestion 7.88E-03 0.705 5.56E-03 total 2.45E+00 N/A 1.47E+00 Default Screening 1.02E+04 (dpm/100 cm 2) N/A 1.70E+04 (dpm/100 cm 2)

Value for 25mrem/y Ag-110m Dose Dose factors Ag-108m Dose (mrem/1.0 pCi/g) (Ag-108m/Ag-110m) (mrem/1.0 pCi/g)

Residential 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 5.12E-22 total 5.14E+00 N/A 3.05E+00 Default Screening 4.87 pCi/g N/A 8.20 pCi/g Value for 25mrem/y 38

January 2012 Appendix D Gross Beta DCGL for Radionuclide Mixture at FNR Appendix A described the mixture of radionuclides, determined for samples from the FNR, during and following completion of remediation. Because of the general absence of contamination in the post-remediation samples, it was decided that the samples, obtained before remediation was completed, would be used to develop a radionuclide mixture for purposes of demonstrating final radiological conditions satisfy the NRC requirements for decommissioning. Table A-2 from Appendix A lists the following contributors to the radionuclide mix:

Concentrations of Mn-54, Zn-65, Ba-133, and Eu-152 were very low in the samples during remediation and were not present in post remediation samples; these radionuclides were therefore deleted from the list of residual radionuclides, leaving only Co-60, Ag-108m, Ag-110m., and Cs-137 as potential contaminants in the remediated facility. The fractional contributions of radionuclides to the pre-remediation and post-remediation mixture, are presented in the following table.

Radionuclide Average Fraction Pre Remediation Post Remediation Co-60 0.312 0.322 Ag-108m 0.588 0.607 Ag-110m 0.048 0.050 Cs-137 0.021 0.021 Mn-54 0.003 Negligible.

Zn-65 0.012 Negligible.

Ba-133 0.012 Negligible.

Eu-152 0.004 Negligible.

Each of the radionuclides in the post-remediation mixture decays to some extent by emitting beta particles. The abundance (A) of beta emissions per decay is 1.0 for Co-60, Ag-110m, and Cs-137 and 0.087 for Ag-108m.

To develop a gross-beta DCGL for the structural surfaces, the fractional contribution (f) of each of the radionuclide contaminants to the total mix was divided by the Default Screening DCGL for that radionuclide. The gross DCGL was then calculated by:

Gross Beta DCGL = fraction of beta emitters (i.e., f x A)

Z (f/DCGL)

Gross Beta DCGL = 0.446 (0.322/7050) + (0.607/17000) + (0.050/10200) + (0.021/28000)

The resulting gross beta DCGL value is 5125 dpm/100 cm 2.

39

January 2012 Appendix E Derived Concentration Guideline Levels for Subsurface Soil at the Former Ford Nuclear Reactor Site University of Michigan Ann Arbor, Michigan Prepared by Denuke Contracting Services, Inc.

Oak Ridge, Tennessee May 4, 2011 40

January 2012 Derived Concentration Guideline Levels for Subsurface Soil at the Former Ford Nuclear Reactor Site

1.0 INTRODUCTION

The Ford Nuclear Reactor (FNR) at the University of Michigan, in Ann Arbor, MI, is being decommissioned. The decommissioning activities are described in University of Michigan DecommissioningPlanfor the FordNuclear Reactor; Chapter 4 of that Decommissioning Plan describes the Final Status Survey Plan for demonstrating that the decommissioning has been effective in satisfying Nuclear Regulatory Commission (NRC) criteria for termination of the license and to permit unrestricted future use of the facility. NRC Default Screening Values have been selected as radiological criteria for the building surfaces and surface soil. However, there are several small regions of potentially contaminated subsurface soil for which Default Screening Values are not applicable. These include soil in the vicinity of the foundation drain around the perimeter of the FNR building; a small volume of soil surrounding a storage tube, used to store several neutron sources; and a small location beneath the freight doorway between the FNR and the adjacent Michigan Memorial Phoenix Project building. The total volume of potentially contaminated subsurface soil is estimated to be less than 50 3

m 3; for calculation purposes, 50 m was the assumed volume. This document describes the development of Derived Concentration Guideline Limits (DCGL's) for radionuclides, that have been identified as potential contaminants in these subsurface soils.

DCGL calculations were based on a dose limit of 25 mrem/y for likely future land use scenarios, RESRAD, Version 6. Calculations were performed, based on deterministic analysis, for a time period out to 1000 years. The most restrictive DCGL for the various scenarios and time periods was selected for each of the potential radionuclide contaminants in the soil.

2.0 POTENTIAL CONTAMINANTS Sampling of accessible soil and structure surfaces in the vicinity of the potentially impacted subsurface soil volumes has identified the following radionuclides as potential contaminants in these soils, resulting from FNR operations:

• Ag-108m

• C-14

• Co-60

• Cs-137

• Eu-152

. 41

January 2012 3.0 SCENARIO DESCRIPTION It is assumed that the University will have institutional control of this site for a period of 100 years, and in the immediate future, after decommissioning is completed, that the former FNR facility will be used by the University primarily as offices, classrooms, and laboratories. After the institutional control period ends, the land could be rezoned for industrial, recreational, or residential use. Three scenarios (A-industrial worker on site, B-building occupancy for educational purposes, and C-off-site resident) were considered for current use conditions. The contamination is limited to a small volume (<50 m 3 ) of subsurface soil, located beneath approximately 5 m of clean cover and about 4 m above the saturated zone; contaminated soil is at least 500 m from the nearest off-site residential area; and surrounding off-site areas are urban and subsistence farming is not a potential land use. The off-site resident scenario was therefore not considered of potential significance and was not evaluated further. Another two scenarios (D-subsistance farming and E- recreational use) were evaluated for the period following institutional control. Because decay of shorter half-life radionuclides would result in lower doses for the industrial use scenario after 100 years of institutional control than for the same scenario during the first 100 years, that scenario was not evaluated beyond the 100 year time period.

Scenario A (current on-site industrial worker) assumes that certain outdoor work activities will result in removing clean overburden, exposing the contaminated soil.

Workers would perform maintenance and construction operations on the surface of the exposed contaminated soil for a period of 3 months, after which the clean overburden would be replaced. The worker would also be exposed annually for 3 months during renovation activities inside the building. The worker will not ingest contaminated water, plant, fish, meat, or milk obtained from the site. This analysis was performed in two steps: Al for exposure to uncovered contaminated soil, and A2 for the work conducted inside the building with the cover in place. The doses from the two steps was then combined for estimating the total industrial worker dose.

Scenario B (building occupant) assumes presence in the building 2000 hr/yr. The occupant would be potentially exposed to external radiation from the contaminated soil.

Exposure to airborne contamination and internal radiation from ingestion of water, plants, meat, milk, and fish, impacted by the contaminated soil are not considered potential exposure pathways.

Scenario D (future subsistence farmer) assumes residential use of the site after the institutional control period. The resident farmer would be exposed to direct radiation and to internal radiation from inhalation of contaminated dust, and ingestion of water, plants, meat, milk, and fish and incidental ingestion of soil. All water used by the 42

January 2012 resident farmer for drinking, irrigation, and household use would be drawn from a deep well, adjacent to the former FNR facility. Nearby surface water sources and shallow wells are not considered adequate to sustain a subsistence farmer.

Scenario E (future recreational use) assumes that an individual camps on the site two weeks per year and engages in recreational activities, such as biking and hiking. During that time drinking water would be drawn from a surface water pond that captures run-off from the site and fish would be obtained from that pond. The recreationist does not ingest plants, meat, or milk obtained from the site.

4.0 PATHWAYS AND KEY PARAMETERS Potential radiation doses resulting from multiple exposure pathways are considered in this analysis for all exposure scenarios. These pathways include:

(1) direct exposure to external radiation from contaminated soil material, (2) internal radiation from inhalation of contaminated dust, (3) internal radiation from incidental ingestion of on-site soil, (4) internal radiation from ingestion of plant foods grown in the contaminated area and irrigated with an on-site water source, (5) internal radiation from ingestion of meat from livestock raised on site and fed with fodder grown in the contaminated area and irrigated with water drawn from an on-site well or pond (the water ingested by livestock is also drawn from an on-site water source),

(6) internal radiation from ingestion of milk from livestock raised onsite and fed with fodder grown in the contaminated area and irrigated with on-site well water or pond water (the water ingested by milk cows is also drawn from on-site water source),

(7) internal radiation from ingestion of fish from a pond located down nearby the contaminated area, (8) internal radiation from drinking water from a deep well, (9) internal radiation from drinking water from an on-site water source (pond water),

(10) internal radiation from ingestion of plant foods grown off-site and irrigated with contaminated well water, 43

January 2012 (11) internal radiation from ingestion of meat from livestock raised off-site and fed with fodder grown off-site but irrigated with contaminated well water (the water ingested by livestock is also drawn from a deep well, (12) internal radiation from ingestion of milk from livestock raised off-site and fed with fodder grown off-site but irrigated with contaminated well water (the water ingested by milk cow is also drawn from the well), and (13) ingestion of fish from an off-site surface water source.

Table I summarizes the applicable exposure pathways for all scenarios and Table 2 lists the key parameters for the applicable exposure pathways. Parameter values for the four scenarios are listed in Table 3. Transfer factors for potential contaminants are presented in Tables 4 and 5. With exception of dimensions and location of the contaminated soil, and a few scenario-specific parameters, RESRAD Default values have been used for these calculations.

Table 1 Summary of Applicable Exposure Pathways for Different Scenarios Considered for Former Ford Nuclear Reactor Facility Exposure Pathway Applicable Pathways Current Use Future Use Scenario Aa Scenario Bb Scenario Dc Scenario Ed Direct external gamma exposure Yes Yes Yes Yes Inhalation of dust Yes* No Yes Yes Ingestion of soil Yes* No Yes Yes Ingestion of plant foods grown onsite No No Yes No Ingestion of meat from livestock raised No No Yes No onsite Ingestion of milk from livestock raised No No Yes No onsite Ingestion of fish from an on-site pond No No Yes Yes Ingestion of water from a downgradient No No Yes No deep well Ingestion of water from a nearby surface No No Yes Yes water source Ingestion of plant foods grown offsite No No No No Ingestion of meat from livestock raised No No No No offsite Ingestion of milk from livestock raised No No No No offsite Ingestion of fish from surface water No No No No source offsite a Industrial Worker b Building Occupant c Subsistence Farmer d Recreationist

• Phase Al only.

44

January 2012 Table 2 List of Key Parameters for Applicable Exposure Pathways Exposure Pathway Key Parameters Direct external gamma exposure Time fraction spent on-site and external gamma shielding factor Inhalation of dust Inhalation rate, time fraction spent on-site, mass loading for inhalation, and indoor dust filtration factor Ingestion of soil Soil ingestion rate and time spent on-site Ingestion of plant foods grown on-site Plant transfer factor, plant ingestion rate, surface water dilution factor, and release rate from the source Ingestion of meat from livestock raised Plant transfer factor, meat transfer factor, meat ingestion rate, water dilution on-site factor, and release rate from the source Ingestion of milk from livestock raised Plant transfer factor, milk transfer factor, milk ingestion rate, surface water on-site dilution factor, and release rate from the source Ingestion of fish from a nearby pond Fish bioaccumulation factor, aquatic food contaminated fraction, surface water dilution factor, and release rate from the source Ingestion of water from a Water ingestion rate, infiltration rate, Kd values (or leach rate) for downgradient well contaminants, and hydrogeological parameters for the site Ingestion of water from a nearby Water ingestion rate, surface water dilution factor, and release rate from the surface water source source Ingestion of plant foods grown off-site Plant transfer factor, plant ingestion rate, surface water dilution factor, and release rate from the source Ingestion of meat from livestock raised Plant transfer factor, meat transfer factor, meat ingestion rate, surface water off-site dilution factor, and release rate from the source Ingestion of milk from livestock raised Plant transfer factor, milk transfer factor, milk ingestion rate, surface water off-site dilution factor, and release rate from the source Ingestion of fish from surface water Fish bioaccumulation factor, aquatic food contaminated fraction, surface source off-site water dilution factor, and release rate from the source 45

January 2012 Table 3. Parameter Values Used for Deterministic Analysis of Different Scenarios _

Current Use Scenarios Future Use Scenarios Input Parameters A: Industrial B: Building D: Subsistence E: Recreationist Comments worker Occupant farmer Title Title Scenario Scenario Scenario Scenario Scenario Definition dependent dependent dependent dependent ICRP-60 methodology based DCFs Dose factor library ICRP-60 ICRP-60 ICRP-60 ICRP-60 (external from FGR-12 and internal from ICRP-72 and FRG-11 (C-14))

Cut-off Half Life (180 d or 30 d) 30 d 30 d 30 d 30 d RESRAD default Number of Points (32, 64, 128, 256, 512, 32 32 32 32 RESRAD default 1024)

Linear Spacing/Log Spacing Log Spacing Log Spacing Log Spacing Log Spacing RESRAD default Maximum No of Points for Dose 17 17 17 17 RESRAD default Maximum No of Points for Risk 1 1 1 1 Use line draw character (yes/no) yes yes yes yes RESRAD default Find peak pathway dose (yes/no) yes yes yes yes Save all files after each run (yes/no) yes yes yes yes Time integrated probabilistic risk no no Dose-to-source ratio (DSR) is calculated (yes/no) no no Calculation Parameters Basic radiation dose limit (mrem/year) 25 25 25 25 Not used in DSR calculation 0,1,,1030,0,1,,1030,Up to the time horizon for dose Times for calculation (years)

____________________________100, 0, 1,3, 10, 30, 300,1000 0, 1, 3, 100, 10, 30, 100, 300, 1000 300,1000caulto 100, 300, 1000 calculation Source Nucide concentration (pCi/g) 1 1 1 1 Dose-source ratios calculated Transport Factors Distribution coefficient for all zones (cm3/g) Table A.4 Table A.4 Table A.4 Table A.4 Number of unsaturated zones 1 1 1 1 RESRAD default Time since placement of material (y) 0 0 0 0 RESRAD default 46

January 2012 Current Use Scenarios Future Use Scenarios Input Parameters A: Industrial B: Building D: Subsistence E: Recreationist Comments worker Occupant farmer Groundwater concentration (pCi/L) 0 0 0 0 RESRAD default Leach rate (l/y) 0 0 0 0 RESRAD default Solubility limit (mol/L) 0 0 0 0 RESRAD default Use plant/soil ration (check box) No No No No RESRAD default Transfer Factors Plant transfer factor (wet-plant weight concentration pCi/g / dry soil weight NAb NA Table A.4 NA concentration pCi/g)

Meat transfer factor (concentration in Table A.4 Table A.4 meat pCi/g / rate of intake pCi/d)

Milk transfer factor (concentration in NA NA Table A.4 NA milk pCi/L / rate of intake pCi/d)

Fish transfer factor (concentration in fish pCi/Kg / concentration in water NA NA Table A.5 TableA5 pCi/L)

Crustacea transfer factor (concentration in Crustacea pCi/Kg / concentration in NA NA Table A.5 Table A.5 water pCi/L)

Contaminated Zone Parameters Area of Contaminated Zone (M 2 ) 25 25 25 25 Characterization survey results Thickness of Contaminated Zone (m) 1 1 1 1 Characterization survey results Length Parallel to Aquifer Flow (m) 100 100 100 100 RESRAD default Does the initial contamination No No No No RESRAD default penetrate the water table?

Cover and Contaminated Zone Hydrological Data A1; 0 Cover Depth (m) A2; 5 5 5 5 Characterization survey results Density of Cover Material (g/cm 3) Al; NA 1.5 1.5 1.5 RESRAD Default

_____ _____ _____ ____ A2; 1.5 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

47

January 2012 Current Use Scenarios Future Use Scenarios Input Parameters A: Industrial B: Building D: Subsistence worker Occupant farmer E: Recreationist Comments Cover Erosion Rate (m/y) 0 0 0 0 RESRAD Default Density of Primary Contaminated Zone 1.5 1.5 1.5 1.5 RESRAD Default (g/cm3 )

Contaminated Zone Erosion Rate 0.001 0.001 0.001 0.001 RESRAD Default (m/y) 0.001_ 0.001 001001ERDefu Contaminated Zone Total Porosity 0.4 0.4 0.4 0.4 RESRAD Default Contaminated Zone Field Capacity 0.2 0.2 0.2 0.2 RESRAD default Contaminated Zone Hydraulic 10 10 10 10 RESRAD Default Conductivity (m/y)

Contaminated Zone b Parameter 5.3 5.3 5.3 5.3 RESRAD Default 3

Humidity in Air (g/m ) NA NA NA NA Used only for H-3 Evapotranspiration Coefficient 0.5 0.5 0.5 0.5 RESRAD Default Wind Speed (m/s) 2.0 2.0 2.0 2.0 RESRAD Default Precipitation rate (m/y) 1.0 1.0 1.0 1.0 RESRAD Default Irrigation rate (m/y) 0.2 NA 0.2 NA RESRAD Default Irrigation mode (Overhead/Ditch) Ooverhead NA Overhead NA RESRAD default Runoff coefficient 0.2 0.2 0.2 0.2 RESRAD Default Watershed area for nearby stream or NA NA 1000000 1000000 RESRAD default pond (M 2 ) NN10100R Ada Accuracy for water/soil computation NA NA 0.001 0.001 RESRAD default Saturated Zone Hydrological Data Density of saturated zone (g/cm3) NA NA 1.5 1.5 RESRAD Default Saturated zone effective porosity NA NA 0.2 0.2 RESRAD Default Saturated zone total porosity NA NA 0.4 0.4 RESRAD Default Saturated zone field capacity NA NA 0.2 0.2 RESRAD Default Saturated zone hydraulic conductivity NA NA 100 100 RESRAD Default (m/y)

Saturated zone b parameter NA NA 5.3 5.3 RESRAD Default 48

January 2012 Current Use Scenarios Future Use Scenarios Input Parameters A: Industrial B: Building D: Subsistence E: Recreationist Comments worker Occupant farmer Saturated zone hydraulic Gradient NA NA 0.02 0.02 RESRAD Default Water table drop rate (m/y) NA NA 0.001 0.001 RESRAD default Well pump intake depth (m below NA NA 10 10 RESRAD Default water table) NANA 10 10 RESRADDefaul Model for water transportation (nondispersion / mass-balance) NA NA Nondispersion Nondispersion RESRAD default Well pumping rate (m 3/y) NA NA 250 250 RESRAD default Unsaturated zone parameters Thickness of unsaturated zone (in) NA NA 10 10 RESRAD Default Density of unsaturated zone (g/cm 3) NA NA 1.5 1.5 RESRAD Default Unsaturated zone effective porosity NA NA 0.2 0.2 RESRAD Default Unsaturated zone total porosity NA NA 0.4 0.4 RESRAD Default Unsaturated zone field capacity NA NA 0.2 0.2 RESRAD default Unsaturated zone hydraulic NA NA 100 100 RESRAD Default conductivity (m/y)

Unsaturated zone b parameter NA NA 5.3 5.3 RESRAD Default Occupancy, Inhalation, and External Gamma Parameters Inhalation rate (m 3/y) 11,400 8400 8400 14,000 Scenario-specific Mass loading for inhalation (g/m 3) 1E-4 1E-4 1E-4 1E-4 RESRAD default Exposure duration (y) 25 30 30 30 Not used in DSR calculations Indoor dust filtration factor NA NA 0.4 0.4 RESRAD default External gamma shielding factor A1;0 0.7 0.7 NA RESRAD default A2; 0.7 0._._ARSAeal Indoor time fraction Al; 0.0833 0.2289 0.6833 0 Scenario-specific A2; 0 Outdoor time fraction Al;0 0 0.0833 0.0384 Scenario-specific A2; 0.0833 Shape of contaminated zone Characterization survey results 25 m x (circular/noncircular) rectangular rectangular rectangular rectangular 2mx 49

January 2012 Current Use Scenarios Future Use Scenarios Input Parameters A: Industrial B: Building D: Subsistence E: Recreatiorust Comments worker Occupant farmer E:_Recrationis Comments lm Ingestion Pathway Dietary Data Fruit, vegetable and grain consumption NA NA 160 NA RESRAD default (kg/y)

Leafy vegetable consumption (kg/y) NA NA 14 NA RESRAD default Milk consumption (L/y) NA NA 92 NA RESRAD default 63 NARESRAD default and scenario-specific NA 63 NArecreationa se Meat and poultry consumption (kg/y) NA for recreational use Fish consumption (kg/y) NA NA 5.4 NA RESRAD frrcetoa default sand scenario-specific for recreational use Other sea food consumption (kg/y) NA NA 0.9 NR RESRAD default.

Soil ingestion (g/y) Al; 36.5 36.5 36.5 36.5 RESRAD default

____ ____ _____ ____ A2; 0 _ _ _ _ _ _ _ _ _ _ _ _ _

Drinking water intake (L/y) NA 510 510 1.4L/d

• 14 = RESRAD default and scenario specific 19.6 L for Recreation use RESRAD default and for offsite Drinking water contaminated fraction NA NA 1 1 receptor scenario water is not contaminated Household water contaminated fractior NA NA 1 NA RESRAD default Livestock water contaminated fraction NA NA 1 1 RESRAD default Irrigation water contaminated fraction NA NA 1 1 RESRAD default Aquatic food contaminated fraction NA NA 0.5 0.5 Distribution from NUREG/CR-6697 Calculated by RESRAD from area Plant food contaminated fraction NA NA -1 NA factor Meat contaminated fraction NA NA -1 NA Scenario specific Milk contaminated fraction NA NA -1 NA Calculated by RESRAD from area factor Ingestion Pathway, Nondietary Data Livestock fodder intake for meat (kg/d) NA NA 68 NA RESRAD default or scenario specific Livestock fodder intake for milk (kg/d) NA NA 55 NA RESRAD default 50

January 2012 Current Use Scenarios Future Use Scenarios Input Parameters A: Industrial B: Building D: Subsistence E: Recreationist Comments worker Occupant farmer Livestock water intake for meat (L/d) NA NA 50 6.4 Or scenario specific Livestock water intake for milk (L/d) NA NA 160 NA RESRAD default Livestock intake of soil for meat (kg/d) NA NA 0.5 NA RESRAD default Livestock intake of soil for milk (kg/d) NA NA 0.5 NA RESRAD default Mass loading for foliar deposition NA NA 0.0001 NA RESRAD default (g/ m3)

Depth of soil mixing layer (m) 0.15 NA 0.15 NA RESRAD default Depth of roots (in) NA NA 0.9 NA RESRAD default Drinking water fraction from 1 NA Scenario specific groundwater Source NNA1 NA Scenario specific Household water fraction from N A1N c n ros eii groundwater Source RESRAD default for Resident Farmer Livestock water fraction from Nand 0 for Recreational Use Scenario groundwater Source because surface water is used by livestock Irrigation water fraction from NA NA 1 NA RESRAD default groundwater Source Plant Factors Wet weight crop yield for non-leafy NA NA 0.7 NA RESRAD default vegetables (kg/m 2)

Length of growing season for non-leafy NA NA 0.17 NA RESRAD default vegetables (y)

Translocation factor for non-leafy NA NA 0.1 NA RESRAD default vegetables Weathering removal constant (l/y) NA NA 20 NA Distribution from NUREG/CR-6697 Wet foliar interception fraction for non- NA NA 0.25 NA RESRAD default leafy vegetables Dry foliar interception fraction or non- NA NA 0.25 NA RESRAD default leafy vegetables I IIIII 51

January 2012 Current Use Scenarios Future Use Scenarios Input Parameters A: Industrial B: Building D: Subsistence E: Recreationist Comments worker Occupant farmer Wet weight crop yield for leafy NA NA 1.5 NA RESRAD default 2

vegetables (kg/m )

Length of growing season for leafy NA NA 0.25 NA RESRAD default vegetables (y)

Translocation factor for leafy vegetables NR NA 1 NA RESRAD default Wet foliar interception fraction for leafy NA NA 0.25 NA RESRAD default vegetables Dry foliar interception fraction for leafy NA NA 0.25 NA RESRAD default vegetables Wet weight crop yield for fodder NA NA 1.1 NA RESRAD default (kg/m2) NANA 1.1_NARESRADdefault Length of growing season for fodder NA NA 0.08 NA RESRAD default (y) NA____ NA_ 0.08____ NA RESRAD______default____

Translocation factor for fodder NA NA 1 NA RESRAD default Wet foliar interception fraction for NA NA 0.25 NA RESRAD default fodder NANA_0.25 _ NA RESRADdefaul Dry foliar interception fraction for NA NA 0.25 NA RESRAD default fodder NNA02NAERDdfu Storage-Times-Before-Use Data Storage time for fruits, non-leafy NA NA 14 NA RESRAD default vegetables and grain (d)

Storage time for leafy vegetables (d) NA NA 1 NA RESRAD default Storage time for milk (d) NA NA 1 NA RESRAD default RESRAD default for Resident Farmer Storage time for meat (d) NA NA 20 NA and scenario specific for Recreational Use RESRAD default and scenario specific Storage time for fish (d) NA NA 7for Recreational use Storage time for crustacea and mollusks RESRAD default and scenario specific (d) NA NA for Recreational use 52

January 2012 Current Use Scenarios Future Use Scenarios A: Industrial B: Building D: Subsistence Input Parameters worker Occupant farmer E: Recreationist Comments Storage time for well water (d) NA NA 1 RESRAD default for Resident Farmer NAannoreuedfrRcatnaUs and not required for Recreational Use Storage time for surface water (d) NA NA 1 NA RESRAD default for Resident Farmer and 0 for Recreational Use RESRAD default for Resident Farmer Storage time for livestock fodder (d) NA NA 45 NA and scenario specific for Recreational Use Radon Datag NA NA NA NA Radon not a .potential contaminant Carbon-14 Datag C-12 concentration in local water, 2E-5 2E-5 2E-5 2E-5 RESRAD default g/cm 3 C-12 concentration in contaminated RESRAD default soil, g/g 0.03 0.03 0.03 0.03 Fraction of vegetation carbon absorbed 0.02 NA 0.02 0.02 RESRAD default from soil Fraction of vegetation carbon absorbed RESRAD default 0.98 0.98 0.98 0.98 from air Thickness of evasion layer of C-14 in 0.3 0.3 0.3 0.3 RESRAD default soil, m C-14 evasion flux rate from soil, sec-1 7E-7 7E-7 7E-7 7E-7 RESRAD default C-12 evasion flux rate from soil, sec-1 1E-10 1E-10 1E-10 1E-10 RESRAD default Grain fraction in beef cattle feed NA NA 0.8 NA RESRAD default Grain fraction in milk cow feed NA NA 0.2 NA RESRAD default 4 .4 4 I I & I NA indicates not applicable to scenario.

53

January 2012 Table 4 Parameter values used in the deterministic analysis for Kd, plant, meat, and milk transfer factors 3

Element Kd (cm /g) Plant transfer Meat transfer Milk transfer Saturated Contaminated and factor factor factor zonea unsaturated zoneb (pCi/kg)/(pCi (pCi/kg)/(pCi/ (pCi/L)/(pCi/d

/ kg) d)

Ag 90 180 1.5 x 10-1 3.0 x 10-3 2.5 x 10-2 C 5 1 7.0 x 10-1 3.1 x 10-2 1.2 x 10-2 Co 60 550 8.0 x 10-2 2.0 x 10-2 2.0 x 10-3 Cs 280 1900 4.0 x 10-2 3.0 X 10-2 8.0 X 10-3 Eu 829 829 2.5 X 10-3 2.0 X 10-3 5.0 x 10-5 a Based on sandy soil type from RESRAD data collection handbook or RESRAD default values for saturated zone b Based on clay soil type from RESRAD data collection handbook or RESRAD default values for contaminated and unsaturated zone Table 5 Parameter Value/Distribution Used in the Analysis for the Fish and Crustacea Transfer Factor Element Crustacea Transfer Factor (concentration Fish Transfer Factor (concentration in in Crustacea pCi/Kg / concentration in Crustacea pCi/Kg / concentration in water pCi/ L) water pCi/L)

Ag 770 5 C 9100 50000 Co 200 300 Cs 100 2000 Eu 1000 50 54

January 2012 5.0 RESULTS OF CALCULATIONS Printouts of computer runs for the 5 potential radionuclide contaminants and 4 exposure scenarios are attached as an appendix to this report (NOTE: computer printouts not included with this appendix to the FSS Plan). Dose estimates for I pCi/g (at t = 0) of soil contamination by each radionuclide of interest are summarized in Tables 6 through 10 for each year of calculation following completion of decommissioning. Values less than 1E-02 mrem/pCi/g would result in DCGL values greater than 2500 p/Ci/g; that value is therefore used as a lower dose bound.

Table 6 Annual Dose from 1 pCi/g of Ag-108m for Evaluated Scenarios Time (yr) Dose (mrem)

Industrial Building Subsistence Recreationist Worker* Occupant Farmer 0 1.919E-01 <1E-02 NA** NA 1 5.376E-02 <1E-02 NA NA 3 <1E-02 <1E-02 NA NA 10 <1E-02 <1E-02 NA NA 30 <1E-02 <1E-02 NA NA 100 <1E-02 <1E-02 <1E-02 <1E-02 300 <1E-02 <1E-02 <1E-02 <1E-02 1000 <1E-02 <1E-02 <1E-02 <1E-02

• Sum of dose from indoor and outdoor activities

• • Scenario not applicable Table 7 Annual Dose from I pCi/g of C-14 for Evaluated Scenarios Time (yr) Dose nmrem)

Industrial Building Subsistence Recreationist Worker* Occupant Farmer 0 <1E-02 <1E-02 NA NA 1 <1E-02 <1E-02 NA NA 3 <1E-02 <1E-02 NA NA 10 <1E-02 <1E-02 NA NA 30 <1E-02 <1E-02 NA NA 100 <1E-02 <1E-02 <1E-02 <1E-02 300 <1E-02 <1E-02 <1E-02 <1E-02 1000 <1E-02 <1E-02 <1E-02 <1E-02

• Sum of dose from indoor and outdoor activities 55

January 2012 Table 8 Annual Dose from 1 pCi/g of Co-60 for Evaluated Scenarios Time (yr) Dose mrem)

Industrial Building Subsistence Recreationist Worker* Occupant Farmer 0 5.357E-01 <1E-02 NA NA 1 4.696E-01 <1E-02 NA NA 3 3.608E-01 <1E-02 NA NA 10 1.434E-01 <1E-02 NA NA 30 1.028E-02 <1E-02 NA NA 100 <1E-02 <1E-02 <1E-02 <1E-02 300 <1E-02 <1E-02 <1E-02 <1E-02 1000 <1E-02 <1E-02 <1E-02 <1E-02

• Sum of dose from indoor and outdoor activities Table 9 Annual Dose from I pCi/g of Cs-137 for Evaluated Scenarios Time (yr) Dose mrem)

Industrial Building Subsistence Recreationist Worker* Occupant Farmer 0 1.182E-01 <1E-02 NA NA 1 1.155E-01 <1E-02 NA NA 3 1.103E-01 <1E-02 NA NA 10 9.375E-02 <1E-02 NA NA 30 5.899E-02 <1E-02 NA NA 100 1.166E-02 <1E-02 <1E-02 <1E-02 300 <1E-02 <1E-02 <1E-02 <1E-02 1000 <1E-02 <1E-02 <1E-02 <1E-02

• Sum of dose from indoor and outdoor activities 56

January 2012 Table 10 Annual Dose from 1 pCi/g of Eu-152 for Evaluated Scenarios Time (yr) Dose (mrem)

Industrial Building Subsistence Recreationist Worker* Occupant Farmer 0 2.413E-01 <1E-02 NA NA 1 2.290E-01 <1E-02 NA NA 3 2.062E-01 <1E-02 NA NA 10 1.430E-01 <1E-02 NA NA 30 5.021E-02 <1E-02 NA NA 100 <1E-02 <1E-02 <1E-02 <1E-02 300 <1E-02 <1E-02 <1E-02 <1E-02 1000 <1E-02 <1E-02 <1E-02 <1E-02

• Sum of dose from indoor and outdoor activities The highest annual dose value for each potential contaminant was chosen as the basis for the subsurface soil DCGL for that radionuclide. Table 11 summarizes the dose values for a concentration of 1 pCi/g at t=O and the resulting DCGL representing a 25 mrem annual dose. These values are the average concentrations, acceptable in the small volume of contaminated soil. Determination of elevated concentrations for smaller volumes was not performed and therefore these values also represent the maximum concentrations acceptable. Implementation of these DCGLs for multiple contaminants will utilize the sum-of-ratios approach.

Table 11 Maximum Annual Dose for I pCi/g Concentration and DCGL Radionuclide Maximum Annual Dose Scenario DCGL (mrem/pCi/g (pCi/g)

Ag-108m 0.1919 Industrial Worker 130 (t=O)

C-14 <lE-02 All (t=-O) 2500 Co-60 0.5357 Industrial Worker 46 (t=O)

Cs-137 0.1182 Industrial Worker 211 (t=0)

Eu-152 0.2413 Industrial Worker 103 (t=O) 57

January 2012 Appendix F DETECTION SENSITIVITIES FOR FNR SURVEY F-I.0 Introduction The final status of the former Ford Nuclear Reactor facility includes surface scans and contamination measurements of structure surfaces and gamma scans of exposed soil surfaces for residual radiological contaminants. Table F-1 is a listing of the radionuclides that have been identified as potential contaminants at the site. In order to assure that residual radioactivity satisfies established decommissioning criteria, instrumentation and survey methods must be capable of detecting residual radioactivity at levels below those criteria. This Appendix describes the determination of radiation detection sensitivities, following the methodology described in NUREG-1507, "Minimum Detectable concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions."

Table F-1 FNR Radiological Contaminants and Decommissioning Criteria (DCGL)

Surface Soil Subsurface Soil Structure Surface DCGL (dpm/100 cm 2)

Radionuclide DCGL (pCi/g) DCGL (pCi/g) Total Removable Co-60 3.8 46 7050 705 Cs-137 11.0 211 28000 2800 Ag-108m 8.2 130 17000 1700 Ag-110m 4.92 N/A 10200 1020 C-14 N/A 2500 N/A N/A Eu-152 N/A 103 N/A N/A Gross beta N/A N/A 5125 512

a. N/A = not applicable or not present as a contaminant.

b. Includes short-lived daughter products present due to assumed ingrowth period of 20 years.

F-2.0 Gamma Scans Gamma scan surveys are conducted using a Ludlum Model 44-10, 2-in.-diameter x 2-in.-thick detector, coupled with a Ludlum Model 2221 scaler/ratemeter, with an open energy window for all photon energies above the input threshold setting of 10 mV. The 58

January 2012 detector is passed over the ground surface in a serpentine pattern. The nominal distance from the detector to the surface is estimated at 5 cm. The serpentine path of the detector is approximately 0.75 to I m in width, and the rate of advancement is approximately 0.5 m/second. The audible signal from the instrument is monitored by the surveyor.

Detectable changes in the count rate are noted, and the immediate area resurveyed at a reduced speed to confirm the change in audible signal and, if applicable, to identify the boundary of the impacted area. The minimum detectable count rate (MDCR) is a function of the background count rate (BKGD) in counts per minute (cpm) and the time (i) in seconds that the detector is within close proximity to the source of radiation.

Equation 6-6 of NUREG-1507 provides the following relationship:

MDCR = d' [BKGD*i/601I/ 2

• 60/i (B1)

For the purposes of application to FNR Final Status Surveys, the area of contaminated soil of concern is assumed to be 0.5 m 2. At a speed of 0.5 m/second, the time (i) in close proximity to the source is, therefore, estimated to be approximately I second. A high probability (95%) of true detection is the objective, and the survey is willing to accept a high probability of false-positive detections (60%) with resulting investigations. The value of d' is selected from Table 6.1 in NUREG-1507 to be 1.38. The nominal site background count rates for the 2 x 2 detectors is 8,000 cpm. Based on these parameters, the MDCR is calculated to be 956 cpm.

The minimum detectable count rate is converted to a radionuclide concentration, using factors representing the detector response in cpm/IpR/h and the exposure rate per unit activity concentration in .IR/h/pCi/g. Values for the detector response for various photon energies are available in tables obtained from NUREG-1507, Table 6.3, and the Multi-Agency Radiation Survey and Assessment of Materialsand Equipment Manual (MARSAME), Appendix F (EPA et al. 2009). A summary of these values is provided in Table F-2. Values for the exposure rate per unit activity were calculated using Version 6.02 of the Microshield computer code. Secular equilibrium of short half-life daughter products with parent was assumed for an ingrowth period of 30 years. The design parameters were uniform concentration in a 50-cm diameter, x 15-cm-thick slab of soil (density 1.5 g/cm3). The exposure rate, was determined at 5 cm above the surface for various photon energies. The products of these two factors for all applicable energy regions was calculated and summed for an overall response factor of cpm/pCi/g associated with a particular radionuclide. These values are summarized in Table F-3.

To account for less than ideal survey performance, a surveyor efficiency factor (p) of (0.5)1/2 was also incorporated into the final calculation as follows:

MDCR (B2)

Scan Sensitivity (pCi/g) = (0.5)1/2 *(cpm/pCi/g)

The resulting scan sensitivity for the potential soil contaminants is also presented in Table F-3, along with the soil DCGL's for comparison.

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January 2012 Table F-2 Detector count rate versus exposure rate (cpm/pR/h)

Energy (keV) 2-in. x 2-in, detector 15 ---

20 2200 30 5160 40 8880 50 11800 60 13000 80 12000 100 9840 150 6040 200 4230 300 2520 400 1700 500 1270 600 1010 800 710 1000 540 1500 350 2000 260 3000 180 Table F-3 Values of cpm/pCi/g and MDC for potential soil contaminants cpm/pCi/g MDC Surface Soil Subsurface Soil Radionuchde (pCi/g) DCGL (pCi/g) DCGL (pCi/g)

Co-60 515 2.62 3.8 46 Cs-137 266 5.08 11.0 211 60

January 2012 Ag-108m 976 1.39 8.2 130 Ag-110m 910 1.49 4.92 N/A C-14 N/Aa N/A N/A 2500 Eu-152 433 3.12 N/A 103 abeta emitter only, not detectable by gamma scans F-3.0 Structure Beta Scans Beta scan surveys are conducted using a Ludlum Model 44-142, 100 cm 2 scintillation detector or a Ludlum Model 43-37, 572 cm 2 gas proportional detector, coupled with a Ludlum Model 2221 scaler/ratemeters. The detector is passed over the surface at a rate of I detector width/sec (i=1) for the Model 44-142 detector and 1/2 detector width / sec (i

= 2) for the Model 43-37 detector, while maintaining the distance from the detector to the surface at approximately 0.5 cm. The audible signal from the instrument is monitored by the surveyor. Detectable changes in the count rate are noted, and the immediate area resurveyed at a reduced speed to confirm the change in audible signal and, if applicable, to identify the boundary of the impacted area. The minimum detectable count rate (MDCR) is a function of the background count rate (BKGD) in counts per minute (cpm) and the time (i) in seconds that the detector is within close proximity to the source of radiation. Equation 6-6 of NUREG-1507 provides the following relationship:

MDCR = d' [BKGD*i/6011/2

• 60/i (B1)

A high probability (95%) of true detection is the objective, and the survey is willing to accept a high probability of false-positive detections (60%) with resulting investigations.

The value of d' is selected from Table 6.1 in NUREG-1507 to be 1.38. The nominal site background count rates for the 100 cm 2 and 572 cm 2 detectors are 300 cpm and 990 cpm, respectively. Based on these parameters, the MDCR is calculated to be 185 cpm for the 100 cm 2 detector and 238 cpm for the 572 cm 2 detector .

Applying detector efficiency values of 0.40 (Model 44-142) and 0.33 (Model 43-37),

surface correction of 0.37, surveyor efficiency of (0.5)1/2, and detector area factors results in an estimated scan detection sensitivity of 1770 dpm/100 cm 2 for the Model 44-142 detector and 480 dpm/100 cm 2 (2760 dpm/total probe area) for the 43-37 detector.

F-4.0 Structure Beta Activity Measurements Direct measurements of beta surface activity are performed using a Ludlum Model 44-142, 100 cm 2 scintillation detector, coupled with a Ludlum Model 2221 scaler/ratemeter.

The detector is placed on the surface and allowed to integrate the count for a period of 1 minute. The MDC is calculated by proximity to the source of radiation. NUREG-1507 provides the following relationship:

MDC = [3 + 4.65 (BKGD)l/ 2 ]/efficiency factors 61

January 2012 The resulting value is 560 dpm/100 cm 2.

Smears for removable activity are counted for 1 minute in a Tennelec low-background alpha/beta counter. The backgrounds are 0.00 alpha cpm and 1.67 beta cpm; 4n detection efficiencies are 0.079 alpha and 0.24 beta. Using the same equation as above for direct measurements yields removable activity MDCs of 38 alpha dpm/100 cm 2 and 38 beta dpm/100 cm 2.

F-5.0. References EPA et al., 2009, Multi-Agency RadiationSurvey andAssessment ofMaterialsand Equipment Manual (MARSAME), EPA 420-R-09-00 1, U.S. Environmental Protection Agency, U.S. Nuclear Regulatory Commission, U.S. Department of Defense, and U.S. Department of Energy, NUREG-1507, 1997, Minimum Detectable Concentrationswith Typical Radiation Survey Instrumentsfor Various Contaminantsand Field Conditions, U.S. Nuclear Regulatory Commission, December 1997. January 2001.

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