Text

TSD-14-021 Revision 0, Bfm Drilling Spoils and Alternate Exposure Scenarios
	ML15344A393
Person / Time
Site:	Zion File:ZionSolutions icon.png
Issue date:	12/18/2014
From:	Farr H, Fauver D, Yetter R; ZionSolutions
To:	; Office of Nuclear Material Safety and Safeguards
Shared Package
ML15344A344	List: ML15344A342; ML15344A355; ML15344A356; ML15344A357; ML15344A358; ML15344A359; ML15344A361; ML15344A362; ML15344A363; ML15344A385; ML15344A386; ML15344A388; ML15344A389; ML15344A390; ML15344A391; ML15344A392; ML15344A393; ML15344A394; ML15344A413; ZS-2013-0082, NPDES Permit for Discharge; ZS-2015-0163, September 24, 2014 Email to Terry Sullivan;
References
	ZS-2015-0163 TSD-14-021, Rev. 0
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Technical Support Document TSD 14-021 BFM Drilling Spoils and Alternate Exposure Scenarios Revision 0 Originator: ___ ______________________________ Date: __12/18/14_______

Harvey Farr Reviewer: __(signature on file)___________________ Date:__12/18/14_______

Dave Fauver Approval: __(signature on file)___________________ Date:__12/18/14_______

Robert F. Yetter

TSD 14-021 Revision 0 Summary of Changes in this Revision:

Rev. 0 -Initial issuance.

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TSD 14-021 Revision 0

1. PURPOSE ...................................................................................................................................... 5

2. DISCUSSION ................................................................................................................................ 5 2.1. End State Basement Fill Model and Resident Farmer Scenario............................ 5 2.2. Alternate Scenarios Evaluated ............................................................................... 6

3. ALTERNATE SCENARIO CALCULATIONS ........................................................................... 7 3.1. BFM Drilling Spoils Scenario ............................................................................... 7 3.2. Inadvertent Intruder Construction Scenario ........................................................ 15 3.3. Large Scale Excavation Scenario ........................................................................ 22

4. WORST CASE CALCULATION TO SUPPORT BFM ELEVATED AREA ASSESSMENT ........................................................................................................................... 32

5. CONCLUSION............................................................................................................................ 35

6. REFERENCES ............................................................................................................................ 36

7. ATTACHMENTS........................................................................................................................ 37 7.1. Attachment A - Construction Scenario MicroShield Reports ............................ 38 7.2. Attachment B - Open Air Demolition MicroShield Reports .............................. 44 LIST OF TABLES Table 1 - Alternate Scenarios Evaluated .................................................................................................. 6 Table 2 - End State Building Well Drilling Parameters........................................................................... 8 Table 3 - Single Nuclide Guidelines and Area Factors Bounding Drill Spoils at 0.15 m Thick ............ 10 Table 4 - End State Building Interpolated Drill Soils Area Factors ....................................................... 11 Table 5 - Drill Spoils Dose Factors mrem/yr per mCi............................................................................ 12 Table 6 - Estimated Drill Spoils Dose for Auxiliary Building and Containment Buildings Estimated End State Bounding Source Terms ........................................................................ 13 Table 7 - Groundwater Pathway Dose for Auxiliary Building and Containment Buildings End State Mixes ...................................................................................................................... 14 Table 8 - Drill Spoils Dose Factors for Radionuclides of Concern ........................................................ 15 Table 9 - Auxiliary Building MicroShield Construction Model Parameters .......................................... 17 Table 10 - Auxiliary Building Basement Inventory Limit and Saturated Zone Activity at t = Peak Concentrations ...................................................................................................................... 18 Table 11 - Calculated Dose Rates and Doses for 5,762 Hour Occupancy In Aux Building Excavation ............................................................................................................................. 19 Table 12 - Spent Fuel Building Basement Fill Inventory Limit and Saturated Zone Activity at t = Peak Concentrations ...................................................................................... 20 Table 13 - Spent Fuel Building MicroShield Construction Model Parameters ...................................... 21 Table 14 - Calculated Dose Rates and Doses for 5,762 Hour Occupancy In Spent Fuel Building Residential Basement............................................................................................................ 22 Table 15 - ROC Maximum Allowable License Termination Source Term ........................................... 22 Table 16 - ROC Inventories Using Aux Building and CTMT Normalized Composite Source Terms .. 23 Table 17 - ROC Max Inventory Fractions for Auxiliary Building and Containment Source Terms ..... 23 Table 18 - Maximum Allowable Inventory at License Termination ...................................................... 24 Table 19 - Decayed Max Inventories at t= Peak Years Post License Termination ................................ 24 Table 20 - Fraction of Activity in Concrete and Sorbed in Fill at t=peak .............................................. 25 Table 21 - Maximum Allowed Activity in Concrete at t=peak ............................................................. 26 Table 22 - Maximum Allowable Activity in Concrete at 50 Years Post License Termination ............. 26 Table 23 - Summary of End State Concrete Volumes and Masses ........................................................ 27 Page 3 of 56

TSD 14-021 Revision 0 Table 24 - Concrete Debris Concentrations at t = 50 years and Soil DCGLs ....................................... 27 Table 25 - Interpolated Area Factors for Concrete Volumes 1 Meter Thick .......................................... 28 Table 26 - Concrete Fractions of Soil DCGLs and Bounding Large Excavation Doses ........................ 28 Table 27 - Maximum Allowed Activity Sorbed in Fill at t=peak ........................................................... 29 Table 28 - Maximum Allowable Activity Sorbed in Fill at 50 Years Post License Termination .......... 29 Table 29 - Excavated Fill Material Volumes and Masses ...................................................................... 30 Table 30 - Fill Concentrations and Soil DCGLs .................................................................................... 30 Table 31 - Interpolated Area Factors for Fill Volumes 1 Meter Thick .................................................. 31 Table 32 - Fill Fractions of Soil DCGLs and Bounding Large Excavation Doses................................. 31 Table 33 - Highest Core Sample and Average Auxiliary Floor Concentration Profiles on July 1, 2018 ........................................................................................................................... 33 Table 34 - July 1, 2018 Estimated Bore Hole Source Terms, Modeled Dose Rates and Open Air Demolition Cut Off Concentrations...................................................................................... 34 Table 35 - July 1, 2018 2a RHR Pump Room Core at 2 mrem/hr Cut Off ............................................ 34 Table 36 - Drill Spoils Concentrations for Aux Building Floor ............................................................. 35 Table 37 - DUST MS Results for Auxiliary Building Diffusion Model with 6503 pCi Source Terms per Nuclide ................................................................................................................ 38 Table 38 - DUST MS Results for Reactor Building Instantaneous Release Model with 2759 pCi Source Terms per Nuclide .................................................................................... 39 Table 39 - DUST MS Results for Spent Fuel Building Diffusion Model with 780 pCi Source Terms per Nuclide ................................................................................................................ 40 Table 40 - DUST MS Results for Turbine Building Instantaneous Release Model with 14,680 pCi Source Terms per Nuclide ................................................................................. 41 Table 41 - DUST MS Results for Crib House/Forebay Building Instantaneous Release Model with 6940 pCi Source Terms per Nuclide ............................................................................ 42 Table 42 - DUST MS Results for Waste Water Treatment Facility Instantaneous Release Model with 1124 pCi Source Terms per Nuclide ............................................................................ 43 TABLE OF FIGURES Figure 1 - Auxiliary Building End State Dimensions ............................................................................. 16 Figure 2 - Top View of Spent Fuel Building End State ......................................................................... 20 Page 4 of 56

TSD 14-021 Revision 0

1. PURPOSE TSD 14-010 (1) and Chapter 6 of the Zion Station Restoration Project (ZSRP) License Termination Plan (LTP) provides the methods for compliance with the radiological criteria for license termination. A Basement Fill Model (BFM) was developed to calculate the dose to the Average Member of the Critical Group (AMCG) from residual radioactivity remaining in the backfilled basements at Zion Nuclear Power Station (ZNPS). The AMCG assumed in the BFM is the Resident Farmer. The BFM conceptual model defines several exposure pathways under an assumption that the configuration of backfilled basements remains in the as-left geometry at the time of license termination. Because the residual radioactivity in the backfilled basements ranges from 15 to 39 feet below grade, depending on the basement, there are no exposure pathways other than through potentially contaminated water from the well that is assumed to be installed as a part of the Resident Farmer exposure scenario.

This technical support document (TSD) evaluates the significance of additional, alternate, exposure pathways resulting from disturbance of the as-left geometry of the backfilled basements. The alternate pathways evaluated include:

exposure to drilling spoils brought to the surface during installation of the resident farmer well, exposure resulting from the construction of a basement to the resident farmer house in the backfill material, and large scale excavation of backfilled structures at some time after license termination.

The drilling spoils exposure scenario is the result of a well being installed into the backfilled basement at the time of maximum BFM groundwater concentration as calculated in TSD 14-010.

(1) This scenario, designated as the BFM Drilling Spoils pathway, was determined to potentially contribute greater than 10% of the total BFM resident farmer dose and was therefore considered significant. The BFM Drilling Spoils pathway was therefore evaluated in detail in this TSD to determine a BFM Drilling Spoils Dose Factor which was included in the BFM dose model for compliance with the license termination criteria. The other two alternate scenarios evaluated in this TSD were determined to not be significant after screening assessments.

2. DISCUSSION 2.1. End State Basement Fill Model and Resident Farmer Scenario As described in the Exelon and ZionSolutions Asset Sale Agreement (2), all structures above the 588 foot elevation will be removed in the site end state. The portions of the structures remaining below the 588 foot elevation will be remediated and surveyed to ensure the 10 CFR 20 Subpart E license termination criteria have been met. They will then be backfilled with clean material.

The results of concrete characterization core samples were evaluated separately for the Reactor Building Containments (3), Auxiliary Building (4) and the Turbine Building. (5). Cores were also obtained in the Crib House, but these were for the purpose of identifying background concentrations in clean, non-contaminated concrete. Only portions of the Spent Fuel Pool below the 588 foot elevation could potentially remain in the end state for the Spent Fuel Building. The rest of the building is above the 588 foot elevation. Since the Spent Fuel Pool is still in use, no characterization Page 5 of 56

TSD 14-021 Revision 0 of end state concrete under the liner has been possible; however, allowable end state source terms that result in 25 millirem per year (mrem/yr) for each radionuclide of concern have been modeled.

(1) The source terms from the concrete characterization data in the Containments and Auxiliary Building are summarized in the Radionuclides of Concern (ROC) TSD 14-019 (6). TSD 14-019 summarizes the calculated source terms from the concrete characterization core decay corrected to the earliest feasible license termination date of July 1, 2018. The TSD also provides the basis for considering the Auxiliary Building as the bounding end state structure relative to potential source term and dose consequences. (4) Since all concrete interior of the Containment liner is to be removed in the Reactor Buildings, the Auxiliary Building and Spent Fuel Pool alternate scenario exposures will bound all other potential doses from end state structures such as the Turbine Building, Waste Water Treatment Facility (WWTF), etc.

The resident farmer scenario, as described in NUREG-1757 Volume 2 Rev. 1 (7), Draft NUREG-1549 (8), and NUREG/CR-6697 (9), assumes a well is installed on a site after license termination and used for drinking, irrigation and livestock water. The resident framer scenario developed for the ZNPS assumes the well is drilled directly into a backfilled basement. The bulk of the Auxiliary Building source term will be in the basement floor at the 542 foot elevation. (4) 2.2. Alternate Scenarios Evaluated Three alternate scenarios are evaluated for the Auxiliary Building End State and the Spent Fuel Pool End State, the Well Installation, Construction and Large Scale Excavation scenarios. The Well Installation Scenario evaluates potential dose from materials brought to the surface at the time of peak fill concentration. The alternate scenarios evaluated are summarized in Table 1.

Table 1 - Alternate Scenarios Evaluated Scenario Description Well Installation Peak fill material concentration Construction Peak fill material concentration Large Scale Excavation After ISFSI Decommissioning The resident farmer scenario assumes a well is placed in the middle of the Auxiliary Building basement or the Spent Fuel Pool structure. This scenario evaluates dose to the resident farmer from soil and material brought to the surface in drilling spoils during the installation process and subsequently dispersed on the surface of the site.

The source term for the drilling spoils scenario is the residual radioactivity in the concrete and fill (after leaching from the surface) at the time of peak water and fill concentrations. The release of source term from the concrete to the fill material is modeled for the full suite of potential ROCs using Brookhaven National Laboratory (BNL) code DUST MS as described in the building models provided in TSD 14-031 (10). The embedded source term in the Auxiliary Building and Spent Fuel Building are assumed to be a diffusion controlled release in very conservative models. The other buildings were modeled as instantaneous release of all concrete source term at t=0. The DUST MS model results for the ROCs determined in TSD 14-019 (6) are provided in Attachment A. As shown in Equation 1, the value in the last column of the Attachment A tables for the Auxiliary and Spent Fuel Building diffusion model is the activity remaining in the concrete at the time of the peak concentration per mCi of original source term.

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TSD 14-021 Revision 0 Equation 1 - Concrete Source Term @ t= Peak per mCi Where:

Cp = Concrete source term conversion factor in pCi per mCi. This is the total source term in all the concrete.

A0 = DUST Model source term in pCi

= Decay Constant in years-1 t = time to peak for ROCs from TSD 14-009 (11)

Aw = Peak Activity in Solution in pCi Af = Peal activity sorbed to full material in pCi A0 mCi = DUST Model source term in mCi NUREG-1757 (7) Volume 2, Appendix J provides an example evaluation for the house construction scenario where radioactive material is brought to the surface during excavation for a house basement. This is commonly referred to as Inadvertent Intruder Construction Scenario. As described in NUREG-1757, the basement excavation is assumed to be about 10 feet deep (3 meters). As noted in the Conestoga-Rovers & Associates (CRA) final hydrology report for ZSRP decommissioning (12), the top grade is at the 591 foot elevation and the top of the water table is at the 579 foot elevation or 12 feet (i.e., 3.66 meters) below the surface. Since the excavation is only to 3 meters, it does not extend into the saturated zone and none of the excavated soil will contain the radioactivity released from the end state structure. However, there is a potential for direct radiation exposure to occupants in the basement from the source term in the saturated zone below the water table. The potential doses for the Construction Scenario at the Auxiliary Building and the Spent Fuel Pool end state areas is assumed to occur at the time of peak source term in basement fill.

3. ALTERNATE SCENARIO CALCULATIONS 3.1. BFM Drilling Spoils Scenario Since the resident farmer scenario assumes a well is installed in the end state basements, the potential exposures from the spoils are evaluated. The scenario assumes a well is drilled with an 8 inch bit and encounters rejection at the floor of the basements after removing a half inch of concrete. The drilling scenario is applied to the Auxiliary Building and Spent Fuel Building which use diffusion controlled release and assumes each radionuclide is at its peak sorbed concentration simultaneously, even though these occur at different times for the ROCs as seen in Attachment A.

The instantaneous release model for the other buildings assumes there is no source term remaining in the concrete and the peak concentration in the fill material occures at t=0. The source term in the fill material below the water table (clean fill contains residual radioactivity as a result of leaching from concrete) and that the source term remaining in the concrete is mixed with the drill spoils from the area above the water table and and spread at a 15 cm depth. Potential exposures are evaluated using site-specific surface soil Derived Concentration Guideline Levels (DCGLs) and area factors derived from RESRAD reports documented in TSD 14-010 (1).

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TSD 14-021 Revision 0 Previous evaluations of exposures from this scenario have assumed a slurry pit was used and that the cuttings were further diluted by refilling the pit with the material excavated during its construction and that it was then later dispersed at the surface. Including a slurry pit results in much lower potential concentrations in the drill spoils spread on the surface. Earthen pits are commonly used adjacent to drilling pits to dispose of drilling mud and well cuttings for oil and gas wells.

Installers vary in how they handle drilling fluids, muds and cuttings differently. Some use earthen pits, some use poly lined pits, some use tanks and the material is sometimes disposed of on-site by dumping it on the surface. As a conservative assumption, no dilution of the end state fill concentrations will be assumed from the drilling mud or in the process of disposing of the slurry.The bore hole will be assumed to be 8 inches in diameter to accommodate the installation of a standard 4 inch diameter casing. The well drilling scenario parameters for each end state building are shown in Table 2.

Table 2 - End State Building Well Drilling Parameters Spent Fuel Crib Rx Bld Pool/ House/

CTMT Transfer Turbine Forebay WWTF Parameter Aux (4) (3) Canals (5) (5) (5) (5)

Ground Surface Elevation ft 591 591 591 591 591 591 Water Table Elevation ft 579 579 579 579 579 579 Basement Floor Elevation ft 542 565 576 560 537 577 Vadose Zone Clean Fill Height ft 12 12 12 12 12 12 Sat Zone Contam Fill Height ft 37 14 3 19 42 2 2

8 inch Diameter Borehole Area (ft ) 0.35 0.35 0.35 0.35 0.35 0.35 Concrete Cutting Depth inches 0.5 0.0 0.5 0.0 0.0 0.0 Concrete Cutting Depth ft 0.042 0.000 0.042 0.000 0.000 0.000 Drill Spoils Volumes Vadose Zone Clean Fill Volume m3 0.119 0.119 0.119 0.119 0.119 0.119 3

Sat Zone Contam Fill Volume m 0.37 0.14 0.03 0.19 0.41 0.02 Density Corrected Concrete Cutting Volume m3 6.59E-04 0.00E+00 6.59E-04 0.00E+00 0.00E+00 0.00E+00 3

Total Borehole Volume m 0.48 0.26 0.15 0.31 0.53 0.14 Spread area m2 @ 0.15 m height 3.23 1.71 0.99 2.04 3.56 0.92 Drilling Spoils Mass Clean Vadose Zone Mass grams 1.78E+05 1.78E+05 1.78E+05 1.78E+05 1.78E+05 1.78E+05 Saturated Zone Contam Bore Hole Mass g 5.48E+05 2.07E+05 4.45E+04 2.82E+05 6.22E+05 2.96E+04 Concrete Cut Mass grams 9.88E+02 0.00E+00 9.88E+02 0.00E+00 0.00E+00 0.00E+00 Total Borehole Mass grams 7.27E+05 3.85E+05 2.23E+05 4.59E+05 8.00E+05 2.07E+05 Saturated Zone Volume and Concrete Floor Areas Sat Zone Volume m3 28445 6537 208 26135 30524 144 3

Modeled Sat Zone Void Space Volume ft 1.00E+06 2.31E+05 7.35E+03 9.23E+05 1.08E+06 5.09E+03 2

Modeled Floor Surface Area (ft ) 27149 16489 2448 48576 25665 2543 Page 8 of 56

TSD 14-021 Revision 0 The grade at the site is at the 591 elevation and the water table is at the 579 elevation. The drill spoils would have 12 feet of clean soil above the water table in the vadose zone. The saturated zone contaminated fill material length of the 8 inch diameter core would be from the floor elevation of each building to the water table at the 579 elevation. Thus the Auxiliary Building would have 12 feet of clean soil from the vadose zone and 37 feet of contaminated soil below it in the saturated zone. The Spent Fuel Building would have 12 feet of clean soil in the vadose zone and 3 feet of contaminated oil in the saturated zone.

A hole diameter of 8 inches equals 0.35 square feet (ft2)which results in the drill spoil volumes and masses shown in Table 2. The drilling spoils volume and mass calculation for the instaneous release buildings (Containments, Turbine, Crib House, and WWTF) take no credit for the mass of the 1/2 inch of concrete cutting drill spoils since the volumes and masses are not significant. The diffusion model buildings (Auxiliary and Spent Fuel) assume that an 8 inch diameter, 1/2 inch deep concrete cutting is removed. The initial density of the concrete is assumed to be 2.40 g/cm3 (3). The concrete drill cutting spoils are asumed to have a density of 1.5 g/cm3, resulting in an increase in the volume by a factor of 1.6. The concrete source terms in the Auxiliary Building and Spent Fuel Building drill spoils are provided in Attachment A Concrete Total at Peak pCi per mCi values in the last column of the tables (e.g. Equation 1 Cp) divided by the floor surface area in Table 2, multiplied by the drill bit surface area of 0.35 ft2.

Equation 2 Diffusion Model Buildings Saturated Zone Drill Spoils Concentration pCi/g per mCi Where Csat zone = The concentration pCi/g per mCi of end state source term in the saturated zone drill spoils from the sorbed activity in the fill and the concrete cuttings.

Cp = The total activity pCi/mCi remaining in the concrete at the peak sorbed fill concentration time (t = peak) calculated using Equation 1.

Af = Table 2 abstracted floor area ft2 of the DUST MS model (e.g. Sat Zone Void Space divided by height of water table above floor).

M sat = The Table 2 mass (grams) of the saturated zone drill spoils.

C sorbed = The DUST MS sorbed concentration pCi/g per mCi from the Attachment A tables.

The estimated concrete cutting activity is very conservative because it assumes all of the activity is concentrated in the first half inch of the floor when core data indictes it is distributed at deeper depths and the inventory calculations included data from wall cores (4) (3). The total 8 inch diameter bore hole volumes in cubic meters (m3) are shown in Table 2. The total drill spoils volumes spread to a 15 cm thickness result in areas of 0.92 to 3.56 m2 as seen in Table 2. The area factors for the full suite of radionuclides are calculated based upon the DCGL(w) reported in the RESRAD single radionuclide guidelines G(i,tmin) calculated for the 64,500 m2 contaminated zone of the site and the 0.3, 1, 3, and 10 m2 contaminated zones. (1) The area factor is calculated by dividing the smaller single nuclide guideline (e.g., 0.3, 1, 3, 10 m2 G(i,tmin)) by the 64,500 m2 G(i,tmin) as provided in Table 3.

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TSD 14-021 Revision 0 Table 3 - Single Nuclide Guidelines and Area Factors Bounding Drill Spoils at 0.15 m Thick RESRAD Report 0.3 m2 1 m2 3 m2 10 m2 0.3 m2 3 m2 10 m2 2 2 64,500 m G(i,tmin) G(i,tmin) G(i,tmin) G(i,tmin) Area 1 m Area Area Area Nuclide pCi/g pCi/g pCi/g pCi/g pCi/g Factors Factors Factors Factors Ag-108m 7.401E+00 2.783E+02 8.349E+01 3.601E+01 1.706E+01 3.76E+01 1.13E+01 4.87E+00 2.31E+00 Am-241 1.872E+02 1.416E+04 7.207E+03 4.132E+03 2.212E+03 7.56E+01 3.85E+01 2.21E+01 1.18E+01 Am-243 5.572E+01 2.176E+03 6.964E+02 3.178E+02 1.539E+02 3.91E+01 1.25E+01 5.70E+00 2.76E+00 C-14 1.196E+02 1.689E+07 4.902E+06 1.710E+06 5.128E+05 1.41E+05 4.10E+04 1.43E+04 4.29E+03 Cm-243 8.561E+01 3.494E+03 1.127E+03 5.064E+02 2.437E+02 4.08E+01 1.32E+01 5.92E+00 2.85E+00 Cm-244 3.966E+02 3.985E+04 3.284E+04 2.565E+04 1.663E+04 1.00E+02 8.28E+01 6.47E+01 4.19E+01 Co-60 4.762E+00 1.947E+02 5.840E+01 2.494E+01 1.177E+01 4.09E+01 1.23E+01 5.24E+00 2.47E+00 Cs-134 7.614E+00 3.366E+02 1.010E+02 4.363E+01 2.068E+01 4.42E+01 1.33E+01 5.73E+00 2.72E+00 Cs-137 1.608E+01 8.012E+02 2.404E+02 1.038E+02 4.920E+01 4.98E+01 1.50E+01 6.46E+00 3.06E+00 Eu-152 1.074E+01 4.168E+02 1.250E+02 5.361E+01 2.534E+01 3.88E+01 1.16E+01 4.99E+00 2.36E+00 Eu-154 9.963E+00 3.917E+02 1.175E+02 5.034E+01 2.378E+01 3.93E+01 1.18E+01 5.05E+00 2.39E+00 Eu-155 3.911E+02 1.140E+04 3.420E+03 1.591E+03 7.783E+02 2.91E+01 8.74E+00 4.07E+00 1.99E+00 Fe-55 3.412E+03 1.166E+09 4.273E+08 1.536E+08 4.754E+07 3.42E+05 1.25E+05 4.50E+04 1.39E+04 H-3 4.817E+03 3.965E+07 1.189E+07 3.964E+06 1.260E+06 8.23E+03 2.47E+03 8.23E+02 2.62E+02 Nb-94 7.510E+00 2.885E+02 8.656E+01 3.726E+01 1.764E+01 3.84E+01 1.15E+01 4.96E+00 2.35E+00 Ni-59 1.175E+04 2.977E+08 9.423E+07 3.198E+07 9.666E+06 2.53E+04 8.02E+03 2.72E+03 8.23E+02 Ni-63 4.289E+03 1.101E+08 3.457E+07 1.170E+07 3.532E+06 2.57E+04 8.06E+03 2.73E+03 8.24E+02 Np-237 8.304E-01 1.268E+03 3.924E+02 1.382E+02 4.578E+01 1.53E+03 4.73E+02 1.66E+02 5.51E+01 Pu-238 2.324E+02 2.491E+04 2.056E+04 1.607E+04 1.042E+04 1.07E+02 8.85E+01 6.91E+01 4.48E+01 Pu-239 2.093E+02 2.265E+04 1.867E+04 1.455E+04 9.390E+03 1.08E+02 8.92E+01 6.95E+01 4.49E+01 Pu-240 2.094E+02 2.268E+04 1.872E+04 1.462E+04 9.462E+03 1.08E+02 8.94E+01 6.98E+01 4.52E+01 Pu-241 9.196E+03 7.451E+05 3.805E+05 2.125E+05 1.118E+05 8.10E+01 4.14E+01 2.31E+01 1.22E+01 Sb-125 3.362E+01 1.259E+03 3.777E+02 1.630E+02 7.726E+01 3.74E+01 1.12E+01 4.85E+00 2.30E+00 Sr-90 2.157E+01 6.204E+04 1.915E+04 6.728E+03 2.215E+03 2.88E+03 8.88E+02 3.12E+02 1.03E+02 Tc-99 1.372E+02 9.305E+05 2.791E+05 9.303E+04 2.956E+04 6.78E+03 2.03E+03 6.78E+02 2.15E+02 The building specific drill spoils area factors are interpolated from the table above using the Table 2 surface area at 0.15 m thick to provide building specific area factors as shown in Table 4 along with the ZSRP 15 cm soil DCGL(w) calculated in TSD 14-010. (1)

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TSD 14-021 Revision 0 Table 4 - End State Building Interpolated Drill Soils Area Factors 2

Drill Spoils m 3.23 1.71 0.99 2.04 3.56 0.92 Crib Nuclide Aux Containment Spent Fuel Turbine House WWTF Ag-108m 4.78 9.00 11.57 7.94 4.66 14.21 Am-241 21.73 32.65 38.91 29.94 21.26 42.64 Am-243 5.61 10.08 12.79 8.96 5.47 15.46 C-14 13966.77 31481.56 42096.81 27087.41 13502.17 52155.67 Cm-243 5.81 10.58 13.47 9.39 5.67 16.25 Cm-244 63.92 76.35 83.00 73.36 62.87 84.77 Co-60 5.15 9.76 12.58 8.60 5.02 15.45 Cs-134 5.63 10.58 13.61 9.34 5.49 16.71 Cs-137 6.34 11.92 15.34 10.53 6.19 18.84 Eu-152 4.90 9.27 11.94 8.18 4.78 14.67 Eu-154 4.96 9.39 12.10 8.28 4.84 14.86 Eu-155 4.00 7.08 8.97 6.31 3.90 11.02 Fe-55 43990.08 96665.85 127632.42 83458.73 42547.34 149359.04 H-3 804.36 1882.34 2532.17 1611.43 778.31 3110.50 Nb-94 4.88 9.19 11.82 8.11 4.75 14.52 Ni-59 2658.93 6132.78 8211.37 5260.52 2570.79 9949.16 Ni-63 2664.96 6161.12 8255.20 5283.20 2576.57 10022.45 Np-237 162.75 363.52 484.22 313.12 157.58 590.04 Pu-238 68.34 81.59 88.68 78.41 67.22 90.55 Pu-239 68.70 82.19 89.41 78.95 67.56 91.32 Pu-240 69.00 82.43 89.61 79.20 67.86 91.51 Pu-241 22.75 34.87 41.82 31.86 22.24 45.79 Sb-125 4.76 8.96 11.52 7.91 4.65 14.16 Sr-90 305.00 682.71 909.83 587.89 295.29 1109.37 Tc-99 662.77 1551.26 2086.84 1327.97 641.30 2563.30 Drill spoils dose factors (mrem/yr per mCi) are calculated from the saturated zone concentrations (Csat zone) calculated in Equation 2 and the Table 4 area factors and DCGLs using equation 3.

Equation 3 - Calculation of Nuclide Specific Drill Spoils Dose Factors mrem/yr per mCi

( )

Where; DF spoils = The mrem/yr per mCi dose factor (DF) of the BFM Drill Spoils pathway.

Csat zone = The concentration (pCi/g per mCi) in the saturated zone from the concrete cutting and the sorbed contamination of the fill material calculated in Equation 2.

Vsat = The volume m3 of the saturated zone core material in Table 2.

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TSD 14-021 Revision 0 Vtotal = The total volume m3 of the core material in Table 2.

AF = The area factor in Table 4.

DCGL(w) = the Soil DCGL pCi/g from Table 4.

25 mrem/yr = The annual dose at the DCGL(w) concentration.

The BFM Drill Spoils dose factors (DFspoils) calculated using Equation 3 are provided in Table 5.

Table 5 - Drill Spoils Dose Factors mrem/yr per mCi Aux Turbine WWTF Building CTMT SFB Drill Drill Crib House Drill Drill Spoils Drill Spoils Spoils Spoils Drill Spoils Spoils mrem/yr mrem per mrem/yr mrem/yr mrem/yr mrem/yr Nuclide per mCi mCi per mCi per mCi per mCi per mCi Ag-108m 1.23E-02 2.05E-02 1.85E-01 6.61E-03 1.23E-02 1.57E-01 Am-241 1.51E-04 3.14E-04 3.06E-03 9.76E-05 1.49E-04 2.91E-03 Am-243 1.58E-03 2.73E-03 2.51E-02 8.75E-04 1.55E-03 2.16E-02 C-14 3.08E-07 4.27E-07 3.72E-06 1.42E-07 3.08E-07 3.11E-06 Cm-243 9.93E-04 1.70E-03 1.56E-02 5.47E-04 9.87E-04 1.34E-02 Cm-244 2.55E-05 6.63E-05 7.09E-04 1.97E-05 2.50E-05 7.21E-04 Co-60 1.07E-02 2.97E-02 1.58E-01 9.58E-03 1.78E-02 2.26E-01 Cs-134 6.29E-03 1.72E-02 9.41E-02 5.54E-03 1.02E-02 1.31E-01 Cs-137 3.22E-03 7.27E-03 4.83E-02 2.35E-03 4.34E-03 5.57E-02 Eu-152 5.02E-03 1.38E-02 7.46E-02 4.45E-03 8.24E-03 1.05E-01 Eu-154 5.57E-03 1.47E-02 8.25E-02 4.73E-03 8.77E-03 1.12E-01 Eu-155 2.83E-04 4.95E-04 4.55E-03 1.58E-04 2.77E-04 3.84E-03 Fe-55 2.97E-09 4.20E-09 3.71E-08 1.39E-09 2.97E-09 3.29E-08 H-3 0.00E+00 0.00E+00 1.45E-09 0.00E+00 0.00E+00 0.00E+00 Nb-94 1.20E-02 1.98E-02 1.79E-01 6.40E-03 1.19E-02 1.52E-01 Ni-59 1.51E-08 2.04E-08 1.77E-07 6.77E-09 1.50E-08 1.52E-07 Ni-63 3.21E-08 5.57E-08 3.75E-07 1.84E-08 4.11E-08 4.13E-07 Np-237 2.91E-03 4.04E-03 3.53E-02 1.34E-03 2.89E-03 3.01E-02 Pm-147 9.32E-08 1.68E-07 1.56E-06 5.35E-08 9.15E-08 1.35E-06 Pu-238 3.97E-05 1.04E-04 1.11E-03 3.08E-05 3.89E-05 1.13E-03 Pu-239 4.41E-05 1.15E-04 1.23E-03 3.40E-05 4.30E-05 1.25E-03 Pu-240 4.38E-05 1.14E-04 1.22E-03 3.38E-05 4.28E-05 1.24E-03 Pu-241 2.97E-06 6.03E-06 5.85E-05 1.88E-06 2.92E-06 5.56E-05 Sb-125 2.75E-03 4.52E-03 4.11E-02 1.46E-03 2.69E-03 3.45E-02 Sr-90 5.84E-05 1.30E-04 7.09E-04 4.31E-05 9.30E-05 9.69E-04 Tc-99 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

= Activated Concrete ROC

= Basement Fill and Soil ROC Page 12 of 56

TSD 14-021 Revision 0 The dose significance of the radionuclides and potential dose contribution from the drill spoils pathway can be assessed using the Table 5 dose factors and the estimated July 1, 2018 Auxiliary Building (4) Unit 1 Containment Buildings (3) estimated activities based 1000 dpm/100 cm2 concrete dust on the liner and using the U1/U2 Containment normalized composite mix from TSD 14-019 (6) derived from the Unit 1 and 2 normalized 568 and Incore characterization data.

Table 6 - Estimated Drill Spoils Dose for Auxiliary Building and Containment Buildings Estimated End State Bounding Source Terms Aux U1 TSD 14-Building Aux CTMT U1 U1 CTMT TSD 14- 019 Drill Profile Drill U1 CTMT CTMT Liner 019 Comp Spoils Drill Spoils CTMT Liner Liner Incore Comp CTMT mrem per Aux Spoils mrem Liner Dust Incore Dust CTMT Mix Nuclide mCi Profile Ci mrem per mCi Dust Ci mrem Dust Ci mrem Mix Ci mrem H-3 0.00E+00 1.46E-03 0.00E+00 0.00E+00 2.92E-05 0.00E+00 2.83E-05 0.00E+00 7.40E-04 0.00E+00 C-14 3.08E-07 3.69E-04 1.14E-07 4.27E-07 5.96E-07 2.54E-10 4.78E-07 2.04E-10 7.71E-05 3.29E-08 Fe-55 2.97E-10 8.68E-04 2.58E-10 4.20E-10 9.94E-07 4.18E-13 8.03E-07 3.37E-13 1.74E-03 7.31E-10 Ni-59 1.51E-08 4.17E-03 6.32E-08 2.04E-08 6.18E-06 1.26E-10 6.01E-06 1.23E-10 1.56E-03 3.19E-08 Co-60 1.07E-02 7.60E-03 8.15E-02 2.97E-02 6.32E-05 1.88E-03 4.74E-05 1.41E-03 4.68E-02 1.39E+00 Ni-63 3.21E-08 1.96E-01 6.30E-06 5.57E-08 6.36E-05 3.54E-09 4.13E-05 2.30E-09 2.63E-01 1.46E-05 Sr-90 5.84E-05 4.27E-04 2.50E-05 1.30E-04 5.91E-07 7.69E-08 8.90E-08 1.16E-08 2.73E-04 3.56E-05 Nb-94 1.20E-02 1.07E-04 1.28E-03 1.98E-02 1.02E-06 2.03E-05 8.70E-07 1.72E-05 1.78E-03 3.53E-02 Tc-99 0.00E+00 1.34E-04 0.00E+00 0.00E+00 3.20E-07 0.00E+00 1.77E-07 0.00E+00 7.59E-05 0.00E+00 Ag-108m 1.23E-02 1.44E-04 1.78E-03 2.05E-02 1.12E-06 2.30E-05 8.42E-07 1.73E-05 2.82E-03 5.79E-02 Sb-125 2.75E-03 1.46E-04 4.03E-04 4.52E-03 5.75E-07 2.60E-06 2.81E-07 1.27E-06 2.48E-04 1.12E-03 Cs-134 6.29E-03 8.60E-05 5.41E-04 1.72E-02 8.22E-07 1.41E-05 7.72E-07 1.33E-05 8.15E-05 1.40E-03 Cs-137 3.22E-03 6.24E-01 2.01E+00 7.27E-03 1.23E-03 8.93E-03 2.86E-04 2.08E-03 6.76E-01 4.91E+00 Eu-152 5.02E-03 1.46E-04 7.31E-04 1.38E-02 4.67E-04 6.43E-03 4.65E-04 6.40E-03 4.36E-03 6.00E-02 Eu-154 5.57E-03 7.89E-05 4.39E-04 1.46E-02 2.24E-05 3.28E-04 2.17E-05 3.18E-04 5.79E-04 8.48E-03 Eu-155 2.83E-04 6.69E-05 1.89E-05 4.95E-04 7.44E-06 3.69E-06 7.27E-06 3.60E-06 1.83E-04 9.05E-05 Np-237 2.91E-03 3.66E-06 1.07E-05 4.04E-03 1.23E-08 4.97E-08 1.09E-08 4.39E-08 9.25E-07 3.74E-06 Pu-238 3.97E-05 1.08E-05 4.29E-07 1.04E-04 2.82E-08 2.93E-09 1.85E-08 1.92E-09 5.19E-06 5.40E-07 Pu-239 4.41E-05 4.47E-06 1.97E-07 1.15E-04 1.21E-08 1.39E-09 8.60E-09 9.85E-10 1.95E-06 2.23E-07 Pu-240 4.38E-05 4.47E-06 1.96E-07 1.14E-04 1.21E-08 1.38E-09 8.59E-09 9.81E-10 1.95E-06 2.22E-07 Pu-241 2.97E-06 2.36E-04 7.03E-07 6.03E-06 7.45E-07 4.49E-09 6.49E-07 3.91E-09 6.70E-05 4.04E-07 Am-241 1.51E-04 1.06E-05 1.61E-06 3.14E-04 1.53E-07 4.81E-08 2.91E-08 9.15E-09 6.73E-05 2.12E-05 Am-243 1.58E-03 7.95E-06 1.26E-05 2.73E-03 1.62E-08 4.42E-08 1.38E-08 3.77E-08 1.71E-06 4.66E-06 Cm-243 9.93E-04 2.80E-06 2.78E-06 1.70E-03 2.17E-08 3.69E-08 1.07E-08 1.82E-08 6.05E-06 1.03E-05 Cm-244 2.55E-05 2.58E-06 6.58E-08 6.63E-05 2.00E-08 1.33E-09 9.82E-09 6.51E-10 5.60E-06 3.71E-07 Total 8.36E-01 2.10E+00 1.89E-03 1.76E-02 9.08E-04 1.03E-02 1.00E+00 6.46E+00 Missed mrem 4.68E-03 4.97E-05 3.95E-05 9.44E-02

% Missed 0.22% 0.28% 0.39% 1.46%

= Activated Concrete

= ROC Bold Italic = MDA ROC Page 13 of 56

TSD 14-021 Revision 0 The Auxiliary Building end state source term is the unremediated estimate for the source term in the concrete floors and walls. The source terms used for the Containment Buildings are those calculated in TSD 13-006 (3) for the liner covered with concrete dust at 1000 dpm/100 cm2. As seen in Table 6, the potential dose from the drill spoils scenario for the Auxiliary Building is significant and in all cases the ROCs contribute over 98% of the dose.

Table 7 - Groundwater Pathway Dose for Auxiliary Building and Containment Buildings End State Mixes U1 U1 TSD 14-CTMT CTMT TSD 14- 019 Aux GW U1 CTMT Liner Liner 019 Comp Aux DF CTMT GW DF Dust Incore Incore Comp CTMT Profile mrem/ Aux GW Liner mrem/ GW Dust GW CTMT Mix GW Nuclide Ci mCi mrem Dust Ci mCi mrem Profile Ci mrem Mix Ci mrem H-3 1.46E-03 6.21E-03 9.05E-03 2.92E-05 2.72E-02 7.92E-04 2.83E-05 7.70E-04 7.40E-04 2.01E-02 C-14 3.69E-04 6.49E-02 2.39E-02 5.96E-07 2.84E-01 1.69E-04 4.78E-07 1.36E-04 7.71E-05 2.19E-02 Fe-55 8.68E-04 8.06E-07 7.00E-07 9.94E-07 1.51E-05 1.50E-08 8.03E-07 1.21E-08 1.74E-03 2.62E-05 Ni-59 4.17E-03 1.35E-04 5.61E-04 6.18E-06 5.87E-04 3.62E-06 6.01E-06 3.53E-06 1.56E-03 9.15E-04 Co-60 7.60E-03 1.00E-04 7.61E-04 6.32E-05 1.14E-02 7.23E-04 4.74E-05 5.42E-04 4.68E-02 5.35E-01 Ni-63 1.96E-01 2.86E-04 5.61E-02 6.36E-05 1.61E-03 1.02E-04 4.13E-05 6.63E-05 2.63E-01 4.22E-01 Sr-90 4.27E-04 3.29E-01 1.41E-01 5.91E-07 4.51E+00 2.67E-03 8.90E-08 4.02E-04 2.73E-04 1.23E+00 Nb-94 1.07E-04 2.03E-03 2.17E-04 1.02E-06 8.84E-03 9.05E-06 8.70E-07 7.68E-06 1.78E-03 1.57E-02 Tc-99 1.34E-04 1.48E-01 1.98E-02 3.20E-07 6.44E-01 2.06E-04 1.77E-07 1.14E-04 7.59E-05 4.89E-02 Ag-108m 1.44E-04 5.47E-03 7.88E-04 1.12E-06 2.56E-02 2.87E-05 8.42E-07 2.16E-05 2.82E-03 7.22E-02 Sb-125 1.46E-04 1.04E-02 1.53E-03 5.75E-07 1.94E-01 1.11E-04 2.81E-07 5.45E-05 2.48E-04 4.81E-02 Cs-134 8.60E-05 9.27E-03 7.97E-04 8.22E-07 1.98E-01 1.62E-04 7.72E-07 1.53E-04 8.15E-05 1.61E-02 Cs-137 6.24E-01 2.64E-02 1.65E+01 1.23E-03 1.57E-01 1.93E-01 2.86E-04 4.48E-02 6.76E-01 1.06E+02 Eu-152 1.46E-04 5.95E-05 8.66E-06 4.67E-04 3.87E-03 1.81E-03 4.65E-04 1.80E-03 4.36E-03 1.69E-02 Eu-154 7.89E-05 6.77E-05 5.34E-06 2.24E-05 5.62E-03 1.26E-04 2.17E-05 1.22E-04 5.79E-04 3.25E-03 Eu-155 6.69E-05 8.01E-06 5.36E-07 7.44E-06 8.72E-04 6.49E-06 7.27E-06 6.34E-06 1.83E-04 1.59E-04 Np-237 3.66E-06 4.92E+01 1.80E-01 1.23E-08 2.13E+02 2.62E-03 1.09E-08 2.32E-03 9.25E-07 1.97E-01 Pu-238 1.08E-05 2.11E-01 2.28E-03 2.82E-08 1.03E+00 2.89E-05 1.85E-08 1.90E-05 5.19E-06 5.33E-03 Pu-239 4.47E-06 2.61E-01 1.17E-03 1.21E-08 1.14E+00 1.38E-05 8.60E-09 9.79E-06 1.95E-06 2.22E-03 Pu-240 4.47E-06 2.61E-01 1.16E-03 1.21E-08 1.14E+00 1.38E-05 8.59E-09 9.79E-06 1.95E-06 2.22E-03 Pu-241 2.36E-04 4.57E-03 1.08E-03 7.45E-07 3.65E-02 2.72E-05 6.49E-07 2.37E-05 6.70E-05 2.45E-03 Am-241 1.06E-05 2.58E-01 2.73E-03 1.53E-07 1.15E+00 1.75E-04 2.91E-08 3.34E-05 6.73E-05 7.72E-02 Am-243 7.95E-06 2.63E-01 2.09E-03 1.62E-08 1.14E+00 1.85E-05 1.38E-08 1.58E-05 1.71E-06 1.95E-03 Cm-243 2.80E-06 2.59E-02 7.26E-05 2.17E-08 1.55E-01 3.37E-06 1.07E-08 1.65E-06 6.05E-06 9.40E-04 Cm-244 2.58E-06 1.74E-02 4.50E-05 2.00E-08 1.24E-01 2.49E-06 9.82E-09 1.22E-06 5.60E-06 6.96E-04 Totals 8.36E-01 1.69E+01 1.89E-03 2.02E-01 5.14E-02 1.09E+02 Spoils Total 2.10E+00 1.76E-02 1.03E-02 6.46E+00 Total Dose mrem/year 1.90E+01 2.20E-01 6.17E-02 1.15E+02 Spoils % 12% 9% 20% 6%

= Activated Concrete ROC = ROC Bold Italic = MDA Page 14 of 56

TSD 14-021 Revision 0 In order to place the above drill spoils doses in context, the groundwater (GW) dose factors from TSD 14-010 (1) can be used to estimate the doses from the groundwater pathway and compared to the drill spoils dose as seen in Table 7. The drill spoils dose contributes 12% of the Auxiliary Building total dose, 9% of the Containment overall mix dose, 20% of the Containment Incore mix which has higher Europium and Tritium fractions and 6% of the Unit 1 and 2 normalized composite mix. The drill spoils are therefoe a significant pathway (i.e., greater than 10% of the 25 mrem/yr dose criterion) that should be included in the evaluation of compliance with the dose criterion. The drill spoils ROCs and their dose factors (DFs) are summarized in Table 8.

Table 8 - Drill Spoils Dose Factors for Radionuclides of Concern Crib Aux Turbine House WWTF Building CTMT SFB Drill Drill Drill Drill Drill Spoils Drill Spoils Spoils Spoils Spoils Spoils mrem/yr mrem per mrem/yr per mrem/yr mrem/yr mrem/yr Nuclide per mCi mCi mCi per mCi per mCi per mCi H-3 0.00E+00 0.00E+00 1.45E-09 0.00E+00 0.00E+00 0.00E+00 Co-60 1.07E-02 2.97E-02 1.58E-01 9.58E-03 1.78E-02 2.26E-01 Ni-63 3.21E-08 5.57E-08 3.75E-07 1.84E-08 4.11E-08 4.13E-07 Sr-90 5.84E-05 1.30E-04 7.09E-04 4.31E-05 9.30E-05 9.69E-04 Cs-134 6.29E-03 1.72E-02 9.41E-02 5.54E-03 1.02E-02 1.31E-01 Cs-137 3.22E-03 7.27E-03 4.83E-02 2.35E-03 4.34E-03 5.57E-02 Eu-152 5.02E-03 1.38E-02 7.46E-02 4.45E-03 8.24E-03 1.05E-01 Eu-154 5.57E-03 1.46E-02 8.25E-02 4.73E-03 8.77E-03 1.12E-01

= Activated Concrete ROC

= Basement Fill and Soil ROC The activated concrete ROCs should be included in addition to the basement fill and soil ROCs for the Containments which will have activated concrete from bioshield and incore area demolition in the residual dust left on the liner. It is unlikely there will be any significant neutron activation associated with the spent fuel since racks include neutron moderation and the storage configuration minimizes neutron flux.

3.2. Inadvertent Intruder Construction Scenario As noted above, the Containment Buildings, Turbine Building, Crib House and Waste Water Treatment Facility (WWTF) end state source terms will be well below the potential activities in the Auxiliary Building and Spent Fuel Building. NUREG-1757 (7) Appendix J provides an example of inadvertent intruder construction which uses a 10 meter by 20 meter (200 m2) house with a 3 meter deep basement. As seen in Table 2, the distance from the surface to the saturated zone is 12 feet.

The basement is 3 meters or 9.8 feet deep, meaning the floor is 2.2 feet above the saturated zone.

Page 15 of 56

TSD 14-021 Revision 0 Figure 1 - Auxiliary Building End State Dimensions As noted in Table 2, the Auxiliary Building basement floor is at the 542 foot elevation and the water table is at the 579 foot elevation.

Sand normally has a dry bulk density of 1.5 g/cc. (13) Site-specific sand bulk densities are higher at 1.81 g/cc (12) and fill material may be concrete at 2.34 g/cc or higher) or a mix of sand and concrete. Since use of a lower density leads to a lower mass of fill material and higher estimated concentration (pCi/g) at the time of maximum concentration as determined in TSD 14-010), this evaluation conservatively used the 1.5 g/cc value for a dry fill mass. At 25% porosity the actual density of the saturated zone source in the MicroShield model used in this assessment would be 1.75 g/cc wet source density as seen in Table 9.

Page 16 of 56

TSD 14-021 Revision 0 Table 9 - Auxiliary Building MicroShield Construction Model Parameters Parameter Value Units Value Units House Basement Length 65.6 ft 20 m House Basement Width 32.8 ft 10 m House Basement Surface Area 2153 ft2 200 m2 Aux Source Length 263 ft 80.1 m Abstracted Aux Source Width 103 ft 31.5 m Aux State Source Height 37 ft 11.3 m Aux Source Volume 2.84E+10 cc 28445.0 m3 Dry Fill Density 1.5 g/cc Dry Source Mass 4.27E+10 grams Wet Source Mass @ 25% Porosity 4.98E+10 grams Wet Source Density 1.75E+00 g/cc Source Activity Co-60 Ci 3.83E-04 Ci 8.98E-03 pCi/g Source Activity Cs-137 Ci 6.04E-01 Ci 1.42E+01 pCi/g Basement Depth 9.8 ft 3 m Dirt Top Shield Distance 2.16 ft 0.66 m Source Density 1.50 g/cc 1.88 g/cc wet Shield Density 1.5 g/cc Dose Point Contact with Excavation 39.199 ft 11.95 m The Auxiliary Building source term and groundwater dose factors in Table 6 and Table 7 can be used to calculate the basement inventory limit equivalent to 25 mrem/yr as shown in Table 10. The total dose at the current estimated inventory is 19.1 mrem/yr, which means the activity could be 1.31 higher at the 25 mrem/yr release criterion. Multiplying the Auxiliary profile activity by the 1.31 correction factor provides the maximum inventory based on the ground water dose factor for the mix. This maximum inventory is actually higher than would be allowed if the drill spoils dose factor was also considered but provides a bounding and conservative saturated zone concentration and source term for the construction scenario. The Auxiliary Building Full DUST MS report (10) data in Attachment A has the total activity in solution (pCi) and sorbed (pCi) at the time of peak fill activity for that nuclide, the sum of these values divided by the DUST MS modeled source term of 6.3503E-06 mCi provides the Table 10 total activity in the fill material conversion factor in pCi/mCi. This factor multiplied by the maximum inventory and divided by 1000 mCi/Ci provides the saturated zone inventory at the time of peak concentration for each nuclide. Since the peak fill concentration times (e.g., t = peak) vary, radioactive decay results in difference in the peak concentration inventories.

Page 17 of 56

TSD 14-021 Revision 0 Table 10 - Auxiliary Building Basement Inventory Limit and Saturated Zone Activity at t = Peak Concentrations Aux Building Saturated Max Sat Zone Drill Zone Total Saturated Total Spoils Aux GW Basement Activity Zone Peak Source @

Aux mrem per DF Aux Dose Inventory pCi per Activity Peak Nuclide Profile Ci mCi mrem/mCi mrem/yr Limit mCi mCi pCi/g Activity Ci H-3 1.46E-03 0.00E+00 6.21E-03 9.05E-03 1.92E+00 1.00E+09 4.50E-02 1.920E-03 C-14 3.69E-04 3.08E-07 6.49E-02 2.39E-02 4.86E-01 1.00E+09 1.14E-02 4.857E-04 Fe-55 8.68E-04 2.97E-10 8.06E-07 7.01E-07 1.14E+00 2.34E+08 6.26E-03 2.669E-04 Ni-59 4.17E-03 1.51E-08 1.35E-04 5.61E-04 5.49E+00 1.00E+09 1.29E-01 5.508E-03 Co-60 7.60E-03 1.07E-02 1.00E-04 8.22E-02 1.00E+01 3.83E+07 8.98E-03 3.831E-04 Ni-63 1.96E-01 3.21E-08 2.86E-04 5.61E-02 2.58E+02 7.79E+08 4.71E+00 2.011E-01 Sr-90 4.27E-04 5.84E-05 3.29E-01 1.41E-01 5.62E-01 3.19E+08 4.20E-03 1.792E-04 Nb-94 1.07E-04 1.20E-02 2.03E-03 1.50E-03 1.40E-01 1.01E+09 3.31E-03 1.410E-04 Tc-99 1.34E-04 0.00E+00 1.48E-01 1.98E-02 1.76E-01 1.00E+09 4.13E-03 1.762E-04 Ag-108m 1.44E-04 1.23E-02 5.47E-03 2.57E-03 1.90E-01 9.37E+08 4.16E-03 1.776E-04 Sb-125 1.46E-04 2.75E-03 1.04E-02 1.93E-03 1.93E-01 2.35E+08 1.06E-03 4.537E-05 Cs-134 8.60E-05 6.29E-03 9.27E-03 1.34E-03 1.13E-01 2.05E+08 5.44E-04 2.321E-05 Cs-137 6.24E-01 3.22E-03 2.64E-02 1.85E+01 8.21E+02 7.36E+08 1.42E+01 6.041E-01 Eu-152 1.46E-04 5.02E-03 5.95E-05 7.39E-04 1.91E-01 6.78E+07 3.04E-04 1.298E-05 Eu-154 7.89E-05 5.57E-03 6.77E-05 4.45E-04 1.04E-01 5.25E+07 1.28E-04 5.454E-06 Eu-155 6.69E-05 2.83E-04 8.01E-06 1.94E-05 8.80E-02 4.00E+07 8.25E-05 3.518E-06 Np-237 3.66E-06 2.91E-03 4.92E+01 1.80E-01 4.81E-03 1.01E+09 1.14E-04 4.846E-06 Pu-238 1.08E-05 3.97E-05 2.11E-01 2.28E-03 1.42E-02 9.00E+08 3.00E-04 1.280E-05 Pu-239 4.47E-06 4.41E-05 2.61E-01 1.17E-03 5.88E-03 1.00E+09 1.38E-04 5.905E-06 Pu-240 4.47E-06 4.38E-05 2.61E-01 1.16E-03 5.87E-03 1.00E+09 1.38E-04 5.895E-06 Pu-241 2.36E-04 2.97E-06 4.57E-03 1.08E-03 3.11E-01 5.49E+08 4.00E-03 1.707E-04 Am-241 1.06E-05 1.51E-04 2.58E-01 2.74E-03 1.40E-02 9.83E+08 3.22E-04 1.373E-05 Am-243 7.95E-06 1.58E-03 2.63E-01 2.10E-03 1.05E-02 1.00E+09 2.46E-04 1.051E-05 Cm-243 2.80E-06 9.93E-04 2.59E-02 7.54E-05 3.69E-03 7.28E+08 6.29E-05 2.685E-06 Cm-244 2.58E-06 2.55E-05 1.74E-02 4.50E-05 3.40E-03 6.11E+08 4.87E-05 2.077E-06 Total 8.36E-01 1.90E+01 1.10E+03 1.91E+01 8.15E-01 Activity CF 1.32

= Activated Concrete ROC = ROC Bold Italic = MDA As seen in Table 9, the abstracted source dimension equivalent to the 262 9.5 length, 37 height and 1.00E+06 cubic feet volume is a 103 foot width instead of the 65 foot width shown for the east area of the Auxiliary Building in Figure 1. This provides a rectangular geometry that can be modeled in MicroShield with the same volume and source term as the overall Auxiliary Building.

The Ni-63, Co-60, and Cs-137 comprise 98.87% of the peak saturated zone source term. Ni-63 is not a gamma emitting nuclide and will not result in exposure through the 2.16 feet of soil above the saturated zone that the foundation would sit on. The Co-60 and Cs-137 source term in Table 10 were used to model the house occupancy potential direct radiation exposures. The dose rates at Page 18 of 56

TSD 14-021 Revision 0 contact (e.g., 1/2 inch) and at 24 inches above the floor were modeled to provide contact and an above the knees whole body dose rate using MicroShield 8.03 using the parameters shown in Table

9. The MicroShield Reports are provided in Attachment B. As a conservative assumption the dry bulk density of soil at 1.5 g/cc was used for the source density even though a wet density based on a very low 25% porosity was used in the MC DUST screening analysis would be 1.75 g/cc for the source. (10)

As seen in the MicroShield report provided in Attachment B, the dose rates in the excavation are very low. The ZSRP resident farmer RESRAD model uses the fraction of time spent indoors of 0.6571 specified in NUREG/CR-5512, Vol. 3 Table 6.87. (14) This equals 5,762 hours0.00882 days 0.212 hours 0.00126 weeks 2.89941e-4 months a year, at the calculated dose rates this equals 0.03 mrem/yr as shown in Table 11.

Table 11 - Calculated Dose Rates and Doses for 5,762 Hour Occupancy In Aux Building Excavation Annual Dose Distance mrem/hr mrem/yr Contact 3.278E-06 1.89E-02 24 inch 5.046E-06 2.91E-02 Given the extremely short construction durations (a few months) and likely actual occupancy times in the basement, as well as the conservative assumptions such as source term dispersal and source density, the potential direct radiation exposure from a construction worker this scenario is insignificant (inhalation and ingestion pathway doses are zero because the source term is 2.2 feet below the excavation floor). The construction scenario does not require consideration as a BFM pathway in demonstrating compliance with the 10 CFR 20 Subpart E license termination requirements.

The Spent Fuel Pool and transfer canal are at the 576 elevation making the saturated zone 3 feet thick. As described in LTP Chapter 6, it is implausible for there to be any groundwater exposure pathways for the Spent Fuel Building end state. Therefore, the drill spoils exposure pathway limits the potential end state source term.

Page 19 of 56

TSD 14-021 Revision 0 Figure 2 - Top View of Spent Fuel Building End State The Auxiliary Building mix can be used to estimate the end state source term using the Spent Fuel Building drill spoils dose factors in Table 5 and the Attachment A DUST MS result as described for the Auxiliary Building above. The peak activity corresponding to the source term limits are shown in Table 12 for the Spent Fuel Building.

Table 12 - Spent Fuel Building Basement Fill Inventory Limit and Saturated Zone Activity at t = Peak Concentrations Saturated MAX Sat Zone Zone Saturated Total SFB Drill Total Zone Source @

Spoils Maximum Activity Peak Peak Aux mrem per SFB Dose Inventory pCi per Activity Activity Nuclide Profile Ci mCi mrem/yr Limit mCi mCi pCi/g Ci H-3 1.46E-03 1.45E-09 2.12E-09 1.16E+00 9.93E+08 3.69E+00 1.153E-03 C-14 3.69E-04 3.72E-06 1.37E-06 2.94E-01 1.00E+09 9.41E-01 2.937E-04 Fe-55 8.68E-04 3.71E-09 3.22E-09 6.91E-01 2.33E+08 5.16E-01 1.610E-04 Ni-59 4.17E-03 1.77E-07 7.38E-07 3.32E+00 1.00E+09 1.06E+01 3.314E-03 Co-60 7.60E-03 1.58E-01 1.20E+00 6.04E+00 3.85E+07 7.45E-01 2.326E-04 Ni-63 1.96E-01 3.75E-07 7.37E-05 1.56E+02 7.78E+08 3.89E+02 1.214E-01 Sr-90 4.27E-04 7.09E-04 3.03E-04 3.40E-01 3.17E+08 3.45E-01 1.076E-04 Nb-94 1.07E-04 1.79E-01 1.91E-02 8.47E-02 9.99E+08 2.71E-01 8.462E-05 Tc-99 1.34E-04 0.00E+00 0.00E+00 1.07E-01 1.00E+09 3.41E-01 1.065E-04 Ag-108m 1.44E-04 1.85E-01 2.67E-02 1.15E-01 9.30E+08 3.42E-01 1.066E-04 Sb-125 1.46E-04 4.11E-02 6.02E-03 1.16E-01 2.34E+08 8.74E-02 2.728E-05 Cs-134 8.60E-05 9.41E-02 8.09E-03 6.84E-02 2.04E+08 4.48E-02 1.396E-05 Cs-137 6.24E-01 4.83E-02 3.02E+01 4.96E+02 7.35E+08 1.17E+03 3.649E-01 Eu-152 1.46E-04 7.46E-02 1.09E-02 1.16E-01 6.67E+07 2.47E-02 7.714E-06 Page 20 of 56

TSD 14-021 Revision 0 Saturated MAX Sat Zone Zone Saturated Total SFB Drill Total Zone Source @

Spoils Maximum Activity Peak Peak Aux mrem per SFB Dose Inventory pCi per Activity Activity Nuclide Profile Ci mCi mrem/yr Limit mCi mCi pCi/g Ci Eu-154 7.89E-05 8.25E-02 6.51E-03 6.28E-02 5.21E+07 1.05E-02 3.272E-06 Eu-155 6.69E-05 4.55E-03 3.04E-04 5.32E-02 3.96E+07 6.74E-03 2.104E-06 Np-237 3.66E-06 3.53E-02 1.29E-04 2.91E-03 9.99E+08 9.31E-03 2.904E-06 Pu-238 1.08E-05 1.11E-03 1.20E-05 8.59E-03 8.92E+08 2.46E-02 7.664E-06 Pu-239 4.47E-06 1.23E-03 5.47E-06 3.55E-03 9.96E+08 1.13E-02 3.539E-06 Pu-240 4.47E-06 1.22E-03 5.46E-06 3.55E-03 9.96E+08 1.13E-02 3.537E-06 Pu-241 2.36E-04 5.85E-05 1.38E-05 1.88E-01 5.46E+08 3.29E-01 1.026E-04 Am-241 1.06E-05 3.06E-03 3.25E-05 8.44E-03 9.78E+08 2.64E-02 8.252E-06 Am-243 7.95E-06 2.51E-02 1.99E-04 6.32E-03 9.99E+08 2.03E-02 6.319E-06 Cm-243 2.80E-06 1.56E-02 4.36E-05 2.23E-03 7.24E+08 5.17E-03 1.612E-06 Cm-244 2.58E-06 7.09E-04 1.83E-06 2.05E-03 6.06E+08 3.99E-03 1.245E-06 Total 8.36E-01 3.14E+01 6.65E+02 1.58E+03 4.92E-01 Activity CF 0.80

= Activated Concrete ROC = ROC Bold Italic = MDA The Spent Fuel Building end state construction scenario modeling parameters are provided in Table

13. The maximum release limit concentrations from the source terms in the saturated zone below the 3 meter deep excavation are 2.33E-04 Ci Co-60 and 3.65E-01 Ci Cs-137 as seen in Table 12.

The Spent Fuel Building MicroShield 8.03 modeling parameters are shown in Table 13.

Table 13 - Spent Fuel Building MicroShield Construction Model Parameters Parameter Value Units Value Units House Basement Length 65.6 ft 20 M House Basement Width 32.8 ft 10 M House Basement Surface Area 2153 ft2 200 m2 Spent Fuel Pool Source Length 63 ft 19.2 M Abstracted Spent Fuel Pool Source Width 39 ft 11.8 M Spent Fuel Pool Source Height 3 ft 0.9 M Spent Fuel Pool Source Volume 2.08E+08 cc 208.0 m3 Dry Fill Density 1.5 g/cc Dry Source Mass 3.12E+08 grams Wet Source Mass @ 25% Porosity 3.64E+08 grams Wet Source Density 1.75E+00 g/cc Source Activity Co-60 Ci 2.33E-04 Ci 7.45E-01 pCi/g Source Activity Cs-137 Ci 3.65E-01 Ci 1.17E+03 pCi/g Basement Depth 9.8 ft 3 m Dirt Top Shield Distance 2.16 ft 0.66 m Page 21 of 56

TSD 14-021 Revision 0 Parameter Value Units Value Units Source Density 1.50 g/cc 1.88 g/cc wet Shield Density 1.5 g/cc Dose Point Contact with Excavation 5.199 ft 1.58 m As seen in Attachment B, the dose rates in the excavation are low. The ZSRP resident farmer RESRAD models uses the fraction of time spent indoors of 0.6571 as specified in NUREG/CR-5512, Vol. 3 Table 6.87. (14) which equals 5,762 hours0.00882 days 0.212 hours 0.00126 weeks 2.89941e-4 months per year. A typical residential floor slab is 4 inches thick with an average concrete density of 2.35 g/cc. The floor slab is included in the Spent Fuel Pool MicroShield model. As seen in the MicroShield report in Attachment B and Table 14 the Construction Scenario dose is 0.45 mrem/yr; thus, any acute exposure during excavation and construction would be trivial.

Table 14 - Calculated Dose Rates and Doses for 5,762 Hour Occupancy In Spent Fuel Building Residential Basement Annual Dose Distance mrem/hr mrem/yr Contact 7.74E-05 4.46E-01 24 inch 7.86E-05 4.53E-01 The estimated doses are less than 2% of the 25 mrem/yr release criterion and are therefore insignificant. The estimated dose is very conservative because it assumes 100% of the indoor occupancy time is spent in the basement and no credit is taken for the side shielding afforded by the concrete footings or vadose zone soil surrounding the basement floor slab. The potential dose from the construction scenario is insignificant and does not require consideration as a BFM pathway in demonstrating compliance with the 10 CFR 20 Subpart E license termination requirements.

3.3. Large Scale Excavation Scenario As a further evaluation of potential alternate scenarios, the concentration of concrete debris and fill material resulting from removal of the end state structures below the 588 foot elevation is examined. The maximum inventory that can remain at license termination and still meet the 25 mrem/year release criteria can be calculated using the Auxiliary Building Mixes and the combined Drill Spoils and Groundwater Dose Factors from TSD 14-010. (1).

The ROC Maximum allowable source term that results in 25 mrem/year for each ROC are shown in Table 15.

Table 15 - ROC Maximum Allowable License Termination Source Term Spent Crib Containment Fuel Turbine House WWTF Nuclide Aux mCi mCi mCi mCi mCi mCi H-3 4.03E+03 9.20E+02 1.72E+10 3.68E+03 4.31E+03 2.02E+01 Co-60 2.31E+03 6.08E+02 1.58E+02 2.01E+03 1.23E+03 3.34E+01 Ni-63 8.75E+04 1.56E+04 6.66E+07 6.23E+04 9.28E+04 3.42E+02 Sr-90 7.59E+01 5.54E+00 3.53E+04 2.21E+01 2.59E+01 1.21E-01 Cs-134 1.61E+03 1.16E+02 2.66E+02 4.55E+02 4.76E+02 2.73E+00 Page 22 of 56

TSD 14-021 Revision 0 Spent Crib Containment Fuel Turbine House WWTF Nuclide Aux mCi mCi mCi mCi mCi mCi Cs-137 8.45E+02 1.52E+02 5.17E+02 6.01E+02 6.60E+02 3.46E+00 Eu-152 4.92E+03 1.42E+03 3.35E+02 4.62E+03 2.76E+03 8.88E+01 Eu-154 4.43E+03 1.23E+03 3.03E+02 4.07E+03 2.51E+03 6.79E+01

= Activated Concrete ROC

= Basement Fill and Soil ROC The normalized composite mixes for the Containments and Auxiliary Building from TSD 14-019 (6) for each radionuclide are applied, are used to calculate ROC inventories for the radionuclide mixes. The ROC mix source terms are provided in Table 16.

Table 16 - ROC Inventories Using Aux Building and CTMT Normalized Composite Source Terms Spent Crib Containment Fuel Turbine House WWTF Nuclide Aux mCi mCi mCi mCi mCi mCi H-3 7.40E-01 Co-60 9.08E+00 4.68E+01 9.08E+00 9.08E+00 9.08E+00 9.08E+00 Ni-63 2.35E+02 2.63E+02 2.35E+02 2.35E+02 2.35E+02 2.35E+02 Sr-90 5.10E-01 2.73E-01 5.10E-01 5.10E-01 5.10E-01 5.10E-01 Cs-134 1.03E-01 8.15E-02 1.03E-01 1.03E-01 1.03E-01 1.03E-01 Cs-137 7.46E+02 6.76E+02 7.46E+02 7.46E+02 7.46E+02 7.46E+02 Eu-152 4.36E+00 Eu-154 5.79E-01

= Activated Concrete ROC

= Basement Fill and Soil ROC The fraction of the Table 15 limit for each source term is shown in Table 17.

Table 17 - ROC Max Inventory Fractions for Auxiliary Building and Containment Source Terms Spent Crib Nuclide Aux Containment Fuel Turbine House WWTF H-3 8.04E-04 Co-60 3.93E-03 7.69E-02 5.75E-02 4.52E-03 7.35E-03 2.71E-01 Ni-63 2.68E-03 1.69E-02 3.53E-06 3.77E-03 2.53E-03 6.87E-01 Sr-90 6.71E-03 4.94E-02 1.45E-05 2.31E-02 1.97E-02 4.20E+00 Cs-134 6.42E-05 7.00E-04 3.88E-04 2.27E-04 2.16E-04 3.78E-02 Cs-137 8.83E-01 4.44E+00 1.44E+00 1.24E+00 1.13E+00 2.15E+02 Eu-152 3.08E-03 Eu-154 4.69E-04 Total 8.96E-01 4.58E+00 1.50E+00 1.27E+00 1.16E+00 2.21E+02

= Activated Concrete ROC

= Basement Fill and Soil ROC Page 23 of 56

TSD 14-021 Revision 0 The Table 16 inventories divided by the sum of the fractions (e.g., Total) in Table 17 calculates the maximum allowed end state source term. The maximum inventory, in mCi, is shown in Table 18.

Table 18 - Maximum Allowable Inventory at License Termination Crib Containment Spent Turbine House WWTF Nuclide Aux mCi mCi Fuel mCi mCi mCi mCi H-3 1.62E-01 Co-60 1.01E+01 1.02E+01 6.05E+00 7.14E+00 7.83E+00 4.11E-02 Ni-63 2.62E+02 5.73E+01 1.57E+02 1.85E+02 2.03E+02 1.06E+00 Sr-90 5.68E-01 5.97E-02 3.40E-01 4.01E-01 4.40E-01 2.31E-03 Cs-134 1.15E-01 1.78E-02 6.88E-02 8.11E-02 8.90E-02 4.67E-04 Cs-137 8.32E+02 1.47E+02 4.97E+02 5.87E+02 6.43E+02 3.38E+00 Eu-152 9.51E-01 Eu-154 1.26E-01 Total 1.10E+03 2.16E+02 6.60E+02 7.79E+02 8.54E+02 4.49E+00

= Activated Concrete ROC

= Basement Fill and Soil ROC The decayed inventories at t = peak years from the DUST MS initial radionuclides of concern model support (10) are calculated by decaying the Table 18 to the t = peak years as shown in Table 19.

Table 19 - Decayed Max Inventories at t= Peak Years Post License Termination Decay Years to t = Containment Spent Fuel Crib House Nuclide Constant Peak Aux mCi mCi mCi Turbine mCi mCi WWTF mCi H-3 5.63E-02 1.00E-01 1.61E-01 Co-60 1.31E-01 4.00E+00 5.98E+00 6.03E+00 3.58E+00 4.22E+00 4.63E+00 2.43E-02 Ni-63 6.92E-03 3.70E+01 2.03E+02 4.44E+01 1.21E+02 1.43E+02 1.57E+02 8.23E-01 Sr-90 2.41E-02 2.10E+01 3.43E-01 3.60E-02 2.05E-01 2.42E-01 2.65E-01 1.39E-03 Cs-134 3.36E-01 1.50E+00 6.95E-02 1.07E-02 4.16E-02 4.90E-02 5.38E-02 2.82E-04 Cs-137 2.30E-02 1.40E+01 6.03E+02 1.07E+02 3.61E+02 4.25E+02 4.66E+02 2.45E+00 Eu-152 5.12E-02 1.00E+01 5.70E-01 Eu-154 7.88E-02 6.00E+00 7.88E-02 Total 8.12E+02 1.58E+02 4.86E+02 5.73E+02 6.28E+02 3.30E+00

= Activated Concrete ROC

= Basement Fill and Soil ROC The DUST MS TSD 14-009 (11) also provides the Peak Activity in Solution in pCi and Peak Activity Sorbed in pCi at t = peak. As seen in Equation 1, the activity in the concrete is the decayed source term modeled in DUST MS for the building and the minus the activity sorbed and activity in solution. Fraction of the source term at t = peak in the concrete is calculated using Equation 4.

Page 24 of 56

TSD 14-021 Revision 0 Equation 4 - Fraction of Source Term in Concrete@ t= Peak per mCi Where:

Fc = Fraction of source term at t=0 in the concrete.

A0 = DUST Model source term in pCi

= Decay Constant in years-1 t = time to peak for ROCs from TSD 14-009 (11)

Aw = Peak Activity in Solution in pCi Af = Peal activity sorbed to full material in pCi A0 mCi = DUST Model source term in mCi Similarly the fraction sorbed in the fill material Equation 5 - Fraction of Source Term Sorbed in Fill @ t= Peak per mCi As seen in Table 20, these fractions are relatively consistent between the Auxiliary Building and the Spent Fuel Pool, the two end state structures to which the diffusion model was applied, despite their extreme variations in concrete surface areas and saturated zone void space volumes. These concrete and fill fractions of source term at t = peak are therefore applicable to the end states of the other structures. The maximum fraction between the two structures was used to ensure conservatism as seen in Table 20.

Table 20 - Fraction of Activity in Concrete and Sorbed in Fill at t=peak Aux SFP Max Max Fraction Fraction Fraction Fraction Fraction Fraction in Sorbed in Sorbed in Sorbed Nuclide Concrete in Fill Concrete in Fill Concrete in Fill H-3 0.00E+00 0.00E+00 1.06E-03 0.00E+00 1.06E-03 0.00E+00 Co-60 9.35E-01 6.48E-02 9.35E-01 6.51E-02 9.35E-01 6.51E-02 Ni-63 0.00E+00 1.00E+00 0.00E+00 1.00E+00 0.00E+00 1.00E+00 Sr-90 4.71E-01 4.93E-01 4.75E-01 4.90E-01 4.75E-01 4.93E-01 Cs-134 6.61E-01 3.38E-01 6.62E-01 3.37E-01 6.62E-01 3.38E-01 Cs-137 0.00E+00 1.00E+00 0.00E+00 1.00E+00 0.00E+00 1.00E+00 Eu-152 8.87E-01 1.13E-01 8.89E-01 1.11E-01 8.89E-01 1.13E-01 Eu-154 9.16E-01 8.41E-02 9.16E-01 8.35E-02 9.16E-01 8.41E-02

= Activated Concrete ROC

= Basement Fill and Soil ROC Page 25 of 56

TSD 14-021 Revision 0 The inventory in the concrete at t = peak is calculated by multiplying the Table 19 activities by the Max Fraction in Concrete from Table 20. This results in the t=0 source terms in the concrete at t =

peak seen in Table 21.

Table 21 - Maximum Allowed Activity in Concrete at t=peak Fraction in Years to t Containment Spent Fuel Turbine Crib WWTF Nuclide Concrete = Peak Aux mCi mCi mCi mCi House mCi mCi H-3 1.06E-03 1.00E-01 1.71E-04 Co-60 9.35E-01 4.00E+00 5.60E+00 5.64E+00 3.34E+00 3.94E+00 4.33E+00 2.27E-02 Ni-63 0.00E+00 3.70E+01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sr-90 4.75E-01 2.10E+01 1.63E-01 1.71E-02 9.74E-02 1.15E-01 1.26E-01 6.62E-04 Cs-134 6.62E-01 1.50E+00 4.60E-02 7.12E-03 2.75E-02 3.25E-02 3.56E-02 1.87E-04 Cs-137 0.00E+00 1.40E+01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Eu-152 8.89E-01 1.00E+01 5.07E-01 Eu-154 9.16E-01 6.00E+00 7.22E-02 Total 5.80E+00 6.24E+00 3.47E+00 4.09E+00 4.49E+00 2.36E-02

= Activated Concrete ROC

= Basement Fill and Soil ROC Given that a large scale excavation is implausible while spent fuel is stored on site, the earliest time such an excavation could occur is at 50 years post license termination. The decay corrected source terms are provided in Table 22.

Table 22 - Maximum Allowable Activity in Concrete at 50 Years Post License Termination Years from Crib t = Peak to Decay Containment Spent Fuel Turbine House WWTF Nuclide t = 50 Constant Aux mCi mCi mCi mCi mCi mCi H-3 4.99E+01 5.63E-02 1.03E-05 Co-60 4.60E+01 1.31E-01 1.32E-02 1.33E-02 7.90E-03 9.31E-03 1.02E-02 5.37E-05 Ni-63 1.30E+01 6.92E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sr-90 2.90E+01 2.41E-02 8.10E-02 8.50E-03 4.84E-02 5.71E-02 6.26E-02 3.29E-04 Cs-134 4.85E+01 3.36E-01 3.91E-09 6.04E-10 2.34E-09 2.76E-09 3.02E-09 1.59E-11 Cs-137 3.60E+01 2.30E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Eu-152 4.00E+01 5.12E-02 6.53E-02 Eu-154 4.40E+01 7.88E-02 2.26E-03 Total 9.42E-02 8.94E-02 5.63E-02 6.64E-02 7.29E-02 3.83E-04

= Activated Concrete ROC

= Basement Fill and Soil ROC The concrete volumes below the 588 foot elevation associated with each building are calculated in TSD 13-006, TSD 14-013, and TSD 14-014 (3), (4), (5). The end state concrete volumes and masses for each building are summarized in Table 23.

Page 26 of 56

TSD 14-021 Revision 0 Table 23 - Summary of End State Concrete Volumes and Masses Total Total Concrete Concrete Concrete Volume per Volume per Mass Structure Item ft3 Item m3 grams Auxiliary Building 5.20E+05 14714.41 3.53E+10 Unit 1 Containment Outside Liner Only 2.21E+05 6269.53 1.50E+10 Unit 2 Containment Outside Liner Only 2.21E+05 6269.53 1.50E+10 Spent Fuel Building 3.94E+04 1116.31 2.68E+09 Turbine Bld, Main Steam Tunnel, Diesel Oil 1.11E+06 31446.15 7.55E+10 Crib House and Forebay 3.46E+05 9788.53 2.35E+10 Waste Water Treatment Facility 1.27E+04 358.88 8.61E+08 Totals 2.47E+06 7.00E+04 1.68E+11 The decayed inventories in Table 22 converted to pCi divided by the concrete masses in Table 23 provide the average concentration of the concrete debris is shown in Table 24 along with the soil DCGLs from TSD 14-010. Soil DCGL concentrations are considered bounding values for screening excavation concrete debris.

Table 24 - Concrete Debris Concentrations at t = 50 years and Soil DCGLs Soil Crib DCGL Containment Spent Fuel Turbine House WWTF Nuclide pCi/g Aux pCi/g pCi/g pCi/g pCi/g pCi/g pCi/g H-3 818.8 1.13E-05 Co-60 3.825 3.74E-04 8.84E-04 2.95E-03 1.23E-04 4.35E-04 6.23E-05 Ni-63 8.486 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sr-90 1.86 2.29E-03 5.65E-04 1.81E-02 7.57E-04 2.67E-03 3.82E-04 Cs-134 4.93 1.11E-10 4.02E-11 8.73E-10 3.65E-11 1.29E-10 1.84E-11 Cs-137 8.606 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Eu-152 9.84 3.37E-02 Eu-154 9.116 4.80E-03 Total 2.67E-03 3.99E-02 2.10E-02 8.80E-04 3.10E-03 4.44E-04

= Activated Concrete ROC

= Basement Fill and Soil ROC Area factors for a 1 meter thick soil are provided in TSD 14-011 (15). The interpolated values that correspond to the Table 23 volumes divided by 1 meter are shown in Table 25.

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TSD 14-021 Revision 0 Table 25 - Interpolated Area Factors for Concrete Volumes 1 Meter Thick Turbine Bld, Main Waste Containment Steam Water Auxiliary Outside Spent Fuel Tunnel, Crib House Treatment Nuclide Building Liner Only Building Diesel Oil and Forebay Facility Co-60 1.06E+00 1.10E+00 1.16E+00 1.04E+00 1.07E+00 1.25E+00 Ni-63 1.73E+00 3.04E+00 6.38E+00 1.49E+00 1.87E+00 2.04E+01 Sr-90 1.27E+00 1.49E+00 1.77E+00 1.18E+00 1.31E+00 5.57E+00 Cs-134 1.20E+00 1.35E+00 1.55E+00 1.13E+00 1.23E+00 1.75E+00 Cs-137 1.29E+00 1.53E+00 1.89E+00 1.19E+00 1.33E+00 2.26E+00 When the Table 24 concentrations are divided by the soil DCGL times the area factor the fractions of the adjusted soil DCGL are calculated. The sum of the fractions of the adjusted soil DCGL times 25 mrem/year provides an estimate of the dose consequence for the concrete excavation. The fraction of the adjusted soil d DCGL and the bounding doses are shown in Table 26.

Table 26 - Concrete Fractions of Soil DCGLs and Bounding Large Excavation Doses Spent Crib Soil Aux Containment Fuel Turbine House WWTF DCGL DCGL DCGL DCGL DCGL DCGL DCGL Nuclide pCi/g Fraction Fraction Fraction Fraction Fraction Fraction H-3 818.8 2.89E-07 Co-60 3.825 9.20E-05 1.50E-04 6.66E-04 1.30E-04 1.06E-04 5.52E-05 Ni-63 8.486 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sr-90 1.86 9.70E-04 2.34E-04 5.49E-03 2.41E-04 1.10E-03 2.61E-05 Cs-134 4.93 1.86E-11 3.23E-11 1.13E-10 3.15E-11 2.12E-11 1.04E-11 Cs-137 8.606 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Eu-152 9.84 1.94E-01 Eu-154 9.116 1.08E-02 Total 1.06E-03 2.05E-01 6.16E-03 3.72E-04 1.20E-03 8.13E-05 Dose mrem/yr 0.03 5.12 0.15 0.01 0.03 0.00

= Activated Concrete ROC

= Basement Fill and Soil ROC These potential worst case doses are well below the potential doses for the groundwater and drill spoils pathways. The alternate scenario of large scale excavation of basement concrete in the future is therefore determined to be insignificant.

The fraction of the source term sorbed in the fill is calculated using Equation 5 from the DUST MS TSD 14-009 (11) data. As seen in Table 20, these fractions are relatively consistent between the Auxiliary Building and the Spent Fuel Pool, the two end state structures to which the diffusion model was applied, despite their extreme variations in concrete surface areas and saturated zone void space volumes. These fill fractions of source term at t = peak are therefore applicable to the end states of the other structures. The maximum fraction between the two structures was used to ensure conservatism as seen in Table 20.

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TSD 14-021 Revision 0 The inventory sorbed in the fill at t = peak is calculated by multiplying the Table 19 activities by the Max Fraction Sorbed in Fill from Table 20. This results in the t=0 source terms in the fill at t = peak seen in Table 27.

Table 27 - Maximum Allowed Activity Sorbed in Fill at t=peak Fraction Spent Crib Sorbed in Years to Containment Fuel Turbine House WWTF Nuclide Fill t = Peak Aux mCi mCi mCi mCi mCi mCi H-3 0.00E+00 1.00E-01 0.00E+00 Co-60 6.51E-02 4.00E+00 3.89E-01 3.92E-01 2.33E-01 2.74E-01 3.01E-01 1.58E-03 Ni-63 1.00E+00 3.70E+01 2.03E+02 4.44E+01 1.21E+02 1.43E+02 1.57E+02 8.23E-01 Sr-90 4.93E-01 2.10E+01 1.69E-01 1.77E-02 1.01E-01 1.19E-01 1.31E-01 6.86E-04 Cs-134 3.38E-01 1.50E+00 2.35E-02 3.63E-03 1.41E-02 1.66E-02 1.82E-02 9.55E-05 Cs-137 1.00E+00 1.40E+01 6.03E+02 1.07E+02 3.61E+02 4.25E+02 4.66E+02 2.45E+00 Eu-152 1.13E-01 1.00E+01 6.44E-02 Eu-154 8.41E-02 6.00E+00 6.62E-03 Total 8.07E+02 1.52E+02 4.82E+02 5.69E+02 6.24E+02 3.28E+00

= Activated Concrete ROC

= Basement Fill and Soil ROC As noted above large scale excavations are not feasible while fuel is stored on site and therefore could not occur for a minimum of 50 years post license termination.

Table 28 - Maximum Allowable Activity Sorbed in Fill at 50 Years Post License Termination Years from t = Crib Peak to t Decay Containment Spent Turbine House WWTF Nuclide = 50 Constant Aux mCi mCi Fuel mCi mCi mCi mCi H-3 4.99E+01 5.63E-02 0.00E+00 Co-60 4.60E+01 1.31E-01 9.19E-04 9.26E-04 5.50E-04 6.48E-04 7.11E-04 3.73E-06 Ni-63 1.30E+01 6.92E-03 1.85E+02 4.06E+01 1.11E+02 1.31E+02 1.43E+02 7.53E-01 Sr-90 2.90E+01 2.41E-02 8.41E-02 8.82E-03 5.03E-02 5.93E-02 6.50E-02 3.41E-04 Cs-134 4.85E+01 3.36E-01 2.00E-09 3.09E-10 1.19E-09 1.41E-09 1.54E-09 8.11E-12 Cs-137 3.60E+01 2.30E-02 2.64E+02 4.67E+01 1.58E+02 1.86E+02 2.04E+02 1.07E+00 Eu-152 4.00E+01 5.12E-02 8.30E-03 Eu-154 4.40E+01 7.88E-02 2.07E-04 Total 4.49E+02 8.73E+01 2.68E+02 3.17E+02 3.47E+02 1.82E+00

= Activated Concrete ROC

= Basement Fill and Soil ROC The estimated fill volume for a large scale excavation is calculated by multiplying the height from the 591 grade elevation to the floor elevation by the abstracted floor surface area in square feet shown in Table 2. The volumes and masses using a density of 1.5 g/cm3 are provided in Table 29.

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TSD 14-021 Revision 0 Table 29 - Excavated Fill Material Volumes and Masses Fill Total Fill Volume Volume 591' to per Item Fill Mass Structure Floor ft3 m3 grams Auxiliary Building 1.33E+06 3.76E+04 5.65E+10 Unit 1 Containment Outside Liner Only 4.29E+05 1.21E+04 1.82E+10 Unit 2 Containment Outside Liner Only 4.29E+05 1.21E+04 1.82E+10 Spent Fuel Building 3.67E+04 1.039E+03 1.56E+09 Turbine Bld, Main Steam Tunnel, Diesel Oil 1.51E+06 4.26E+04 6.40E+10 Crib House and Forebay 1.39E+06 3.92E+04 5.89E+10 Waste Water Treatment Facility 3.56E+04 1.01E+03 1.51E+09 Totals 1.68E+11 1.68E+11 1.68E+11 The decay corrected sorbed source terms at 50 years divided by the Table 29 masses provide the maximum fill material concentrations in pCi/g that could theoretically exist after license termination.

Table 30 - Fill Concentrations and Soil DCGLs Soil Spent Crib DCGL Aux Containment Fuel Turbine House WWTF Nuclide pCi/g pCi/g pCi/g pCi/g pCi/g pCi/g pCi/g H-3 818.8 0.00E+00 Co-60 3.825 1.63E-05 5.09E-05 3.52E-04 1.01E-05 1.21E-05 2.47E-06 Ni-63 8.486 3.28E+00 2.23E+00 7.10E+01 2.04E+00 2.43E+00 4.98E-01 Sr-90 1.86 1.49E-03 4.84E-04 3.22E-02 9.26E-04 1.10E-03 2.26E-04 Cs-134 4.93 3.53E-11 1.69E-11 7.65E-10 2.20E-11 2.62E-11 5.37E-12 Cs-137 8.606 4.67E+00 2.57E+00 1.01E+02 2.91E+00 3.46E+00 7.09E-01 Eu-152 9.84 4.56E-04 Eu-154 9.116 1.14E-05 Total 7.95E+00 4.79E+00 1.72E+02 4.95E+00 5.90E+00 1.21E+00

= Activated Concrete ROC

= Basement Fill and Soil ROC Area factors for a 1 meter thick soil are provided in TSD 14-011 (15). The interpolated values that correspond to the Table 29 volumes divided by 1 meter are shown in Table 31.

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TSD 14-021 Revision 0 Table 31 - Interpolated Area Factors for Fill Volumes 1 Meter Thick Turbine Bld, Main Waste Containment Steam Water Auxiliary Outside Spent Fuel Tunnel, Crib House Treatment Nuclide Building Liner Only Building Diesel Oil and Forebay Facility Co-60 1.03E+00 1.05E+00 1.16E+00 1.03E+00 1.03E+00 1.16E+00 Ni-63 1.39E+00 1.09E+00 6.47E+00 1.32E+00 1.37E+00 6.51E+00 Sr-90 1.15E+00 1.19E+00 1.78E+00 1.12E+00 1.14E+00 1.78E+00 Cs-134 1.11E+00 1.15E+00 1.56E+00 1.09E+00 1.10E+00 1.56E+00 Cs-137 1.16E+00 1.20E+00 1.90E+00 1.13E+00 1.15E+00 1.90E+00 When the Table 30 concentrations are divided by the soil DCGL times the area factor the fractions of the adjusted soil DCGL are calculated. The sum of the fractions of the adjusted soil DCGL times 25 mrem/year provides an estimate of the dose consequence for the concrete excavation. The fraction of the adjusted soil DCGL and the bounding doses are shown in Table 32.

Table 32 - Fill Fractions of Soil DCGLs and Bounding Large Excavation Doses Spent Aux Containment Fuel Turbine Crib House WWTF Soil DCGL DCGL DCGL DCGL DCGL DCGL DCGL Nuclide pCi/g Fraction Fraction Fraction Fraction Fraction Fraction H-3 818.8 0.00E+00 Co-60 3.825 4.11E-06 1.26E-05 7.94E-05 2.58E-06 3.06E-06 5.57E-07 Ni-63 8.486 2.77E-01 2.40E-01 1.29E+00 1.82E-01 2.09E-01 9.01E-03 Sr-90 1.86 6.97E-04 2.18E-04 9.74E-03 4.45E-04 5.21E-04 6.82E-05 Cs-134 4.93 6.47E-12 3.00E-12 9.96E-11 4.10E-12 4.83E-12 6.98E-13 Cs-137 8.606 4.69E-01 2.49E-01 6.19E+00 2.99E-01 3.51E-01 4.34E-02 Eu-152 9.84 4.63E-05 Eu-154 9.116 1.25E-06 Total 7.46E-01 4.89E-01 7.50E+00 4.82E-01 5.60E-01 5.24E-02 Dose mrem/yr 18.66 12.23 187.39 12.05 14.01 1.31

= Activated Concrete ROC

= Basement Fill and Soil ROC As seen in Table 32, the potential dose from the fill excavation scenario could bound the drilling spoils dose for the spent fuel pool. All other doses are less than the 25 mrem/year based upon the maximum allowable source term that could be left and still meet the release criteria. It should be noted that these are maximum allowable source term and that the source terms in the all the end state structures are significantly less with the exception of the Spent Fuel Pool and Transfer Canal which remain to be characterized.

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TSD 14-021 Revision 0

4. WORST CASE CALCULATION TO SUPPORT BFM ELEVATED AREA ASSESSMENT A bounding calculation was performed to support an assessment in LTP Chapter 6 regarding the potential impact of the hypothetical worst-case radionuclide concentrations that could remain in the Auxiliary basement (and possibly the Spent Fuel Pool although characterization has not yet been performed) after demolition to the 2 mR/hr open air demolition limits. (16) To risk-inform the acceptability of the worst-case concentrations, the potential dose consequences of this worst-case concentrations are evaluated. The dose was assessed using the drilling spoils scenario described in Section 3.1 with the exception that the highest concentrations that could hypothetically remain in the Auxiliary Basement after remediation to the open air demolition limits are used as the concrete source term. This scenario is called the Worst-Case Drilling Spoils.

As described in LTP Chapter 6, the Worst-Case Drilling Spoils assessment is considered a less likely but plausible scenario (as defined in NUREG-1757, Table 5.1). Consistent with NUREG 1757, Table 5.1 the scenario is not analyzed as an alternate scenario but is used to help risk inform and justify the decision that the hypothetical maximum concentrations that could remain in elevated areas after remediation to the 2 mR/hr demolition limit are acceptable assuming all activity is accounted for by the BFM inventory using the ground water and drill spoils dose factors.

The less likely but plausible Worst-Case Drilling Spoils scenario assumes that the water supply well is drilled directly into a spot of residual radioactivity with the highest hypothetical concentration immediately after license termination taking no credit for decay or release to the fill water. The entire inventory in the spot is assumed to be excavated and brought to the surface while mixing with overburden fill and soil. This is very unlikely for two reasons. First, the scenario assumes that a Resident Farmer water supply well is installed immediately after license termination, while the Independent Spent Fuel Storage Installation (ISFSI) is present, which is a highly unlikely, essentially non-credible, land use (as discussed in section 6.5.3). Second, the probability of an assumed eight inch borehole hitting an area containing the maximum hypothetical contamination level during drilling is low. For example, the area in the Auxiliary Basement floor with the highest contamination levels is limited to ~20 m2 (in two RHR rooms) of the ~2500 m2 total floor area.

Note that the dose from the worst-case drilling spoil scenario is separate and distinct from the BFM dose in that it is assumed to occur before any release of activity from the concrete and therefore the water and fill concentrations are zero.

This scenario assumes an inadvertent intruder drills into the first half inch of a concrete floor at t=0.

As seen in Table 33, the highest sample from the Auxiliary Building floor was from the 2A Residual Heat Removal Pump Room. (4). The concentrations of the 2A RHR and average core decay corrected to July 1, 2018 are shown in Table 33. The 2A RHR Pump core sample levels are approximately an order of magnitude higher than the overall average concentrations in the first few inches of concrete. However, the depth profile is similar with the majority of the source term in the first two inches.

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TSD 14-021 Revision 0 Table 33 - Highest Core Sample and Average Auxiliary Floor Concentration Profiles on July 1, 2018 First 2 Inches 86% 96% 81% 83%

AUX 542 2A RHR Average all Aux AUX 542 2A RHR Average all Floor Pump Floor Cores Pump Cores B105103- B105103- B105103- B105103-Top CJFCCV- CJFCCV- Avg. Avg. CJFCCV- CJFCCV- Avg. Avg.

Depth 001 Co-60 001 Cs- Co-60 Cs-137 001 Co-60 001 Cs- Co-60 Cs-137 Puck # inches pCi/g 137 pCi/g pCi/g pCi/g % 137 % % %

Puck 1 0 196.1 20205.4 20.3 2751.2 26% 47% 28% 41%

Puck 2 0.5 172.4 11353.3 12.0 1266.7 23% 27% 16% 19%

Puck 3 1 124.5 5430.6 11.6 737.6 16% 13% 16% 11%

Puck 4 1.5 79.0 2210.2 7.7 450.9 10% 5% 10% 7%

Puck 5 2 83.1 1782.9 7.7 364.1 11% 4% 11% 5%

Puck 6 2.5 47.7 1187.1 5.3 260.5 6% 3% 7% 4%

Puck 7 3 30.0 319.9 3.4 196.8 4% 1% 5% 3%

Puck 8 3.5 25.3 195.1 2.8 234.7 3% 0% 4% 4%

Puck 9 4 NS NS 0.2 114.8 0.0% 0% 0.3% 2%

Puck 10 4.5 NS NS 0.1 286.6 0.0% 0% 0.2% 4%

Notes: NS equals Not Sampled As noted in TSD 10-002 (16), there are limits for open air demolition of concrete that include a less than 2 millirem per hour (mrem/hr) on contact requirement. The activity of a 1 foot diameter source is calculated by multiplying by the concentration in pCi/g times the 2224 gram mass of 1/2 inch thick section with a density of 2.35 g/cc and dividing by 1E+12 pCi/Ci. MicroShield 8.03 was used to calculate the dose rate each core depth would contribute for a one foot diameter area at the 2A RHR Pump core concentrations. The reports are provided in Attachment B and are summarized in Table

34. As seen in Table 34, the contact dose rate on the 2A RHR Pump core would be approximately 3.3 mrem/hr. This is slightly above the 2 mrem/hr contact limit; thus, this area would be remediated.

In addition, past the two inch depth within the concrete, the source term contributes less than 1% of the contact dose rates due to the lower concentration and shielding of the concrete layers above. The concentrations at 61% of those in the 2A RHR Pump core puck sample would equal the 2 mrem/hr open air demolition limit.

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TSD 14-021 Revision 0 Table 34 - July 1, 2018 Estimated Bore Hole Source Terms, Modeled Dose Rates and Open Air Demolition Cut Off Concentrations 2A RHR Max Max Pump 2A RHR Cut- Cut-One Pump Dose Off Off Top Foot Dia One Foot Rate at Percent Conc Conc Depth Co-60 Dia Cs- Surface Dose Co-60 Cs-137 Sample inches Ci 137 Ci mrem/hr Rate pCi/g pCi/g Puck 1 0 1.90E-07 1.95E-05 2.06 62% 118.8 12234.2 Puck 2 0.5 1.67E-07 1.10E-05 0.77 23% 104.5 6874.3 Puck 3 1 1.21E-07 5.25E-06 0.26 8% 75.5 3288.2 Puck 4 1.5 7.64E-08 2.14E-06 0.11 3% 47.9 1338.3 Puck 5 2 8.04E-08 1.72E-06 0.07 2% 50.3 1079.5 Puck 6 2.5 4.62E-08 1.15E-06 0.03 1% 28.9 718.8 Cut-Off 2 mrem/hr Total 3.3 CF 0.61 Adjusting the upper 2.5 inches to the cut off value concentrations in Table 34, the source term of the remediated profile in the 8 inch diameter drill is shown in Table 35.

Table 35 - July 1, 2018 2a RHR Pump Room Core at 2 mrem/hr Cut Off B105103- B105103- 2A RHR 2A RHR Top CJFCCV- CJFCCV- Pump 8 Pump 8 Depth 001 Co-60 001 Cs- Inch Dia Inch Dia Sample inches pCi/g 137 pCi/g Co-60 pCi Cs-137 pCi Puck 1 0 118.8 12234.2 1.15E+05 1.18E+07 Puck 2 0.5 104.5 6874.3 1.01E+05 6.65E+06 Puck 3 1 75.5 3288.2 7.30E+04 3.18E+06 Puck 4 1.5 47.9 1338.3 4.63E+04 1.30E+06 Puck 5 2 50.3 1079.5 4.87E+04 1.04E+06 Puck 6 2.5 28.9 718.8 2.80E+04 6.96E+05 As noted in Table 2, there are 7.27E+05 grams in the Auxiliary Building drill spoils including the 1/2 inch concrete cutting and there are 2.23E+05 grams in the Spent Fuel Building drill spoils. The Cs-137 scaling factors of the July 1, 2018 source terms are calculated from the Auxiliary Building concrete mix in Table 10. The source terms of the ROCs are calculated from the first half inch puck Co-60 and Cs-137 activities in Table 35. When compared to the area factor adjusted soil DCGLs in Table 36, the fraction of the Auxiliary Building DCGL is 0.169. Thus, the potential dose from a rejected drilling attempt at the hypothetical worst-case concentration that could remain after remediation to the open air demotion limit is 5.68 mrem/yr for the Auxiliary Building, and 0.227 or 5.08 mrem/yr for the Spent Fuel Building.

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TSD 14-021 Revision 0 Table 36 - Drill Spoils Concentrations for Aux Building Floor Drill Aux Area SFB Area Cutting Aux Drill SFB Drill Factor Aux Factor SFB Cs-137 Source Spoil Spoil DCGL DCGL DCGL DCGL Nuclide SF Term pCi pCi/g pCi/g pCi/g Fraction pCi/g Fraction H-3 0.26% 3.08E+04 4.23E-02 1.38E-01 3.68E+06 1.15E-08 1.16E+07 1.19E-08 Co-60 1.09% 1.15E+05 1.58E-01 5.15E-01 2.44E+01 6.49E-03 5.96E+01 8.65E-03 Ni-63 31.41% 3.72E+06 5.12E+00 1.67E+01 1.06E+07 4.81E-07 3.30E+07 5.05E-07 Sr-90 0.04% 5.05E+03 6.94E-03 2.26E-02 4.39E+03 1.58E-06 1.31E+04 1.73E-06 Cs-134 0.01% 1.20E+03 1.66E-03 5.40E-03 4.24E+01 3.91E-05 1.02E+02 5.27E-05 Cs-137 100.00% 1.18E+07 1.63E+01 5.31E+01 1.00E+02 1.63E-01 2.42E+02 2.20E-01 Eu-152 0.02% 2.48E+03 3.41E-03 1.11E-02 5.27E+01 6.47E-05 1.28E+02 8.66E-05 Eu-154 0.01% 1.38E+03 1.90E-03 6.17E-03 4.95E+01 3.83E-05 1.21E+02 5.12E-05 Total 1.33 1.57E+07 21.63 1.70E-01 2.28E-01 mrem 4.24E+00 5.71E+00

5. CONCLUSION Based upon this evaluation, the inadvertent intruder construction alternate scenario will not result in significant exposure and will not require consideration for demonstrating compliance with the 10 CFR 20 Subpart E (17) license termination criteria. However, the BFM Drill Spoils scenario has the potential to result in exposures that are significant relative to the license termination criteria (i.e.,

exceeding 10% of the 25 mrem/yr dose criteria) and were therefore evaluated in detail and the BFM Drilling Spoils pathway dose added to the BFM for demonstrating compliance with the 10 CFR 20 Subpart E (17) license termination criteria. Table 8 provides the BFM Drilling Spoils Dose Factors.

The large scale excavation scenario could result in significant exposure for the Spent Fuel Pool and Transfer Canal end states and should be re-evaluated when characterization data and estimated source term inventories are available.

The dose from a hypothetical worst-case concentration in concrete after remediation to the open air demolition limits was calculated for use in LTP Chapter 6 to risk-inform the evaluation of elevated areas in the BFM.

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TSD 14-021 Revision 0

6. REFERENCES

1. TSD 14-010, RESRAD Dose Modeling for Basement Fill Model and Soil DCGL and Calculation of Basement Fill Model Dose Factors.

2. Zion Nuclear Power Station, Units 1 And 2 Asset Sale Agreement, December 11, 2007.

3. TSD 13-006 Reactor Building Units 1 and 2 End State Concrete and Liner Initial Characterization Source Terms and Distributions.

4. TSD 14-013 Zion Auxiliary Building End State Estimated Concrete Volumes, Surface Areas, and Source Terms.

5. TSD 14-014 End State Surface Areas, Volumes, and Source Terms of Ancillary Buildings.

6. TSD 14-019 Radionuclides of Concern for Soil and Basement Fill Model Source Terms.

7. NUREG-1757 Vol. 2, Rev. 1, Consolidated Decommissioning Guidance Characterization, Survey, and Determination of Radiological Criteria, September 2006.

8. Draft NUREG-1549, Decision Methods for Dose Assessment to Comply With Radiological Criteria for License Termination, July 1998.

9. NUREG/CR-6697, Development of Probabilistic RESRAD 6.0 and RESRAD-BUILD 3.0 Computer Codes, U.S. Nuclear Regulatory Commission, December 2000.

10. TSD 14-031 BNL Report: Basement Fill Model Evaluation of Maximum Radionuclide Concentrations for Initial Suite of Radionuclides.

11. TSD 14-009 BNL Report: Evaluation of Maximum Radionuclide Groundwater Concentrations for Basement Fill Model.

12. Evaluation of Hydrological Parameters in Support of Dose Modeling for the Zion Restoration Project, Conestoga-Rovers & Associates, Chicago, IL, January 14, 2014, Reference No.054638, Revision 4, Report No. 3.

13. ANL/EAIS-8, Data Collection Handbook to Support Modeling the Impacts of Radioactive Material in Soil, Yu, C., et. al, Argonne National Laboratory, Argonne, IL, 1993.

14. NUREG/CR 5512 Vol. 3 Residual Radioactive Contamination From Decommissioning Parameter Analysis, October 1999 http://pbadupws.nrc.gov/docs/ML0824/ML082460902.pdf .

15. TSD 14-011 Soil Area Factors.

16. TSD 10-002 Technical Basis for Radiological Limits for Structure, Building Open Air Demolition.

17. 10 CFR 20 Standards for Protection Against Radiation, Subpart ERadiological Criteria for License Termination.

18. TSD 14-005 Backfill Material Specifications.

19. TSD 14-015, Buried Piping Dose Modeling and Derived Concentration Guideline Levels.

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TSD 14-021 Revision 0

7. ATTACHMENTS 7.1. Attachment A - Construction Scenario MicroShield Reports 7.2. Attachment B - Open Air Demolition MicroShield Reports Page 37 of 56

Attachment A TSD 14-021 DUST MS Full Radionuclide Suite Results Revision 0 Table 37 - DUST MS Results for Auxiliary Building Diffusion Model with 6503 pCi Source Terms per Nuclide Peak Time Activity Peak Peak Sorbed Concrete Decay Diffusion to Peak in Activity Sorbed Conc Total at Constant Kd Coefficient Peak Conc Solution Sorbed Conc pCi/g per Peak pCi Nuclide yr-1 (ml/g) (cm2/s) (years) pCi/L pCi pCi pCi/g mCi per mCi H-3 5.63E-02 0 5.50E-07 0.1 9.10E-04 6503 0 0.00E+00 0.000E+00 0.000E+00 C-14 1.22E-04 1.2 3.00E-09 1.11E-04 793 5710 1.33E-07 2.045E-02 0.000E+00 Fe-55 2.53E-01 2857 3.00E-09 1.24E-08 0.09 1519 3.54E-08 5.444E-03 7.664E+08 Ni-59 6.86E-06 62 1.10E-09 2.45E-06 17.4 6512 1.52E-07 2.337E-02 0.000E+00 Co-60 1.31E-01 223 4.10E-11 4 2.60E-08 0.2 249 5.80E-09 8.919E-04 5.527E+08 Ni-63 6.92E-03 62 1.10E-09 37 1.90E-06 13.6 5051 1.18E-07 1.815E-02 0.000E+00 Sr-90 2.41E-02 2.3 5.20E-10 21 1.96E-05 140.1 1933 4.51E-08 6.935E-03 2.844E+08 Nb-94 3.41E-05 45 3.00E-09 3.38E-06 24 6521 1.52E-07 2.337E-02 0.000E+00 Tc-99 3.28E-06 0 5.50E-07 9.15E-04 6503 0 0.00E+00 0.000E+00 0.000E+00 Ag-108m 1.66E-03 27 3.00E-09 5.23E-06 37 6054 1.41E-07 2.168E-02 6.336E+07 Sb-125 2.51E-01 17 3.00E-09 2.08E-06 15 1516 3.54E-08 5.444E-03 7.646E+08 Cs-134 3.36E-01 45 3.00E-09 1.5 6.89E-07 5 1329 3.10E-08 4.767E-03 3.992E+08 Cs-137 2.30E-02 45 3.00E-09 14 2.47E-06 17.7 4766 1.11E-07 1.707E-02 0.000E+00 Pm-147 2.64E-01 95 3.00E-09 3.68E-07 3 1499 3.50E-08 5.382E-03 7.690E+08 Eu-152 5.12E-02 95 5.00E-11 10 1.07E-07 0.8 440 1.02E-08 1.569E-03 5.315E+08 Eu-154 7.88E-02 95 5.00E-11 6 8.38E-08 0.6 341 7.96E-09 1.224E-03 5.708E+08 Eu-155 1.46E-01 95 5.00E-11 6.39E-08 0 260 6.07E-09 9.334E-04 9.600E+08 Np-237 3.23E-07 1 3.00E-09 1.31E-04 936 5616 1.31E-07 2.014E-02 0.000E+00 Pu-238 7.90E-03 174 3.00E-09 7.84E-07 6 5848 1.36E-07 2.091E-02 9.980E+07 Pu-239 2.87E-05 174 3.00E-09 8.75E-07 6 6527 1.52E-07 2.337E-02 0.000E+00 Pu-240 1.06E-04 174 3.00E-09 8.74E-07 6 6519 1.52E-07 2.337E-02 0.000E+00 Pu-241 4.83E-02 174 3.00E-09 4.78E-07 3 3566 8.32E-08 1.279E-02 4.512E+08 Am-241 1.60E-03 177 3.00E-09 8.42E-07 6 6389 1.49E-07 2.291E-02 1.661E+07 Am-243 9.40E-05 177 3.00E-09 8.60E-07 6 6526 1.52E-07 2.337E-02 0.000E+00 Cm-243 2.43E-02 889 3.00E-09 1.24E-07 1 4736 1.10E-07 1.692E-02 2.716E+08 Cm-244 3.83E-02 889 3.00E-09 1.04E-07 1 3973 9.27E-08 1.425E-02 3.889E+08 Note: Concrete Total at Peak pCi per mCi in last column is total activity remaining in concrete at time to peak using Equation 1.

Page 38 of 56

Attachment A TSD 14-021 DUST MS Full Radionuclide Suite Results Revision 0 Table 38 - DUST MS Results for Reactor Building Instantaneous Release Model with 2759 pCi Source Terms per Nuclide Peak Peak Sorbed Decay Activity in Activity Peak Conc Constant Peak Conc Solution Sorbed Sorbed pCi/g per Nuclide yr-1 Kd (ml/g) pCi/L pCi pCi Conc pCi/g mCi H-3 5.63E-02 0 1.69E-03 2759 0 0.00E+00 0.000E+00 C-14 1.22E-04 1.2 2.06E-04 336.6 2422.4 2.47E-07 8.953E-02 Fe-55 2.53E-01 2857 9.82E-08 0.2 2758.8 2.81E-07 1.018E-01 Ni-59 6.86E-06 62 4.53E-06 7.4 2751.6 2.81E-07 1.018E-01 Co-60 1.31E-01 223 1.26E-06 2.1 2756.9 2.81E-07 1.018E-01 Ni-63 6.92E-03 62 4.53E-06 7.4 2751.6 2.81E-07 1.018E-01 Sr-90 2.41E-02 2.3 1.14E-04 186.4 2572.6 2.62E-07 9.496E-02 Nb-94 3.41E-05 45 6.23E-06 10.2 2748.8 2.80E-07 1.015E-01 Tc-99 3.28E-06 0 1.69E-03 2759 0 0.00E+00 0.000E+00 Ag-108m 1.66E-03 27 1.04E-05 16.9 2742.1 2.80E-07 1.015E-01 Sb-125 2.51E-01 17 1.64E-05 26.7 2732.3 2.79E-07 1.011E-01 Cs-134 3.36E-01 45 6.23E-06 10.2 2748.8 2.80E-07 1.015E-01 Cs-137 2.30E-02 45 6.23E-06 10.2 2748.8 2.80E-07 1.015E-01 Pm-147 2.64E-01 95 2.95E-06 4.8 2754.2 2.81E-07 1.018E-01 Eu-152 5.12E-02 95 2.95E-06 4.8 2754.2 2.81E-07 1.018E-01 Eu-154 7.88E-02 95 2.95E-06 4.8 2754.2 2.81E-07 1.018E-01 Eu-155 1.46E-01 95 2.95E-06 4.8 2754.2 2.81E-07 1.018E-01 Np-237 3.23E-07 1 2.41E-04 394 2365 2.41E-07 8.735E-02 Pu-238 7.90E-03 174 1.62E-06 2.6 2756.4 2.81E-07 1.018E-01 Pu-239 2.87E-05 174 1.62E-06 2.6 2756.4 2.81E-07 1.018E-01 Pu-240 1.06E-04 174 1.62E-06 2.6 2756.4 2.81E-07 1.018E-01 Pu-241 4.83E-02 174 1.62E-06 2.6 2756.4 2.81E-07 1.018E-01 Am-241 1.60E-03 177 1.59E-06 2.6 2756.4 2.81E-07 1.018E-01 Am-243 9.40E-05 177 1.59E-06 2.6 2756.4 2.81E-07 1.018E-01 Cm-243 2.43E-02 891 3.15E-07 0.5 2758.5 2.81E-07 1.018E-01 Cm-244 3.83E-02 891 3.15E-07 0.5 2758.5 2.81E-07 1.018E-01 Page 39 of 56

Attachment A TSD 14-021 DUST MS Full Radionuclide Suite Results Revision 0 Table 39 - DUST MS Results for Spent Fuel Building Diffusion Model with 780 pCi Source Terms per Nuclide Peak Activity Peak Peak Sorbed Concrete Decay Diffusion Time to Peak in Activity Sorbed Conc Total at Constant Kd Coefficient Peak Conc Solution Sorbed Conc pCi/g per Peak pCi Nuclide yr-1 (ml/g) (cm2/s) (years) pCi/L pCi pCi pCi/g mCi per mCi H-3 5.63E-02 0 5.50E-07 0.1 1.49E-02 774.8 0 0.00E+00 0.000E+00 1.056E+06 C-14 1.22E-04 1.2 3.00E-09 1.83E-03 95.2 685.2 2.20E-06 2.821E+00 0.000E+00 Fe-55 2.53E-01 2857 3.00E-09 2.04E-07 0.011 181.8 5.83E-07 7.474E-01 7.669E+08 Ni-59 6.86E-06 62 1.10E-09 4.02E-05 2.1 777.6 2.49E-06 3.192E+00 3.846E+05 Co-60 1.31E-01 223 4.10E-11 4 4.25E-07 0.022 30 9.48E-08 1.215E-01 5.525E+08 Ni-63 6.92E-03 62 1.10E-09 37 3.13E-05 1.6 605 1.94E-06 2.487E+00 0.000E+00 Sr-90 2.41E-02 2.3 5.20E-10 21 3.21E-04 16.7 230.3 7.38E-07 9.462E-01 2.865E+08 Nb-94 3.41E-05 45 3.00E-09 5.53E-05 2.9 776.4 2.49E-06 3.192E+00 8.974E+05 Tc-99 3.28E-06 0 5.50E-07 1.50E-02 780 0 0.00E+00 0.000E+00 0.000E+00 Ag-108m 1.66E-03 27 3.00E-09 8.56E-05 4.5 721.1 2.31E-06 2.962E+00 6.974E+07 Sb-125 2.51E-01 17 3.00E-09 3.41E-05 1.8 180.9 5.80E-07 7.436E-01 7.658E+08 Cs-134 3.36E-01 45 3.00E-09 1.5 1.13E-05 0.6 158.7 5.09E-07 6.526E-01 4.002E+08 Cs-137 2.30E-02 45 3.00E-09 14 4.07E-05 2.1 571.4 1.83E-06 2.346E+00 0.000E+00 Pm-147 2.64E-01 95 3.00E-09 6.03E-06 0.3 178.7 5.73E-07 7.346E-01 7.705E+08 Eu-152 5.12E-02 95 5.00E-11 10 1.75E-06 0.09 51.9 1.66E-07 2.128E-01 5.326E+08 Eu-154 7.88E-02 95 5.00E-11 6 1.37E-06 0.07 40.6 1.30E-07 1.667E-01 5.712E+08 Eu-155 1.46E-01 95 5.00E-11 1.04E-06 0.05 30.8 9.88E-08 1.267E-01 9.604E+08 Np-237 3.23E-07 1 3.00E-09 2.14E-03 111.3 667.7 2.14E-06 2.744E+00 1.282E+06 Pu-238 7.90E-03 174 3.00E-09 1.28E-05 0.67 694.9 2.23E-06 2.859E+00 1.082E+08 Pu-239 2.87E-05 174 3.00E-09 1.43E-05 0.74 776.3 2.49E-06 3.192E+00 3.795E+06 Pu-240 1.06E-04 174 3.00E-09 1.43E-05 0.74 776.3 2.49E-06 3.192E+00 3.795E+06 Pu-241 4.83E-02 174 3.00E-09 7.83E-06 0.41 425.1 1.36E-06 1.744E+00 4.545E+08 Am-241 1.60E-03 177 3.00E-09 1.38E-05 0.72 762.1 2.44E-06 3.128E+00 2.203E+07 Am-243 9.40E-05 177 3.00E-09 1.41E-05 0.73 778.7 2.50E-06 3.205E+00 7.308E+05 Cm-243 2.43E-02 889 3.00E-09 2.03E-06 0.11 564.3 1.81E-06 2.321E+00 2.764E+08 Cm-244 3.83E-02 889 3.00E-09 1.70E-06 0.09 472.6 1.51E-06 1.936E+00 3.940E+08 Note: Concrete Total at Peak pCi per mCi in last column is total activity remaining in concrete at time to peak using Equation 1 Page 40 of 56

Attachment A TSD 14-021 DUST MS Full Radionuclide Suite Results Revision 0 Table 40 - DUST MS Results for Turbine Building Instantaneous Release Model with 14,680 pCi Source Terms per Nuclide Peak Peak Decay Activity in Activity Peak Sorbed Constant Peak Conc Solution Sorbed Sorbed Conc pCi/g Nuclide yr-1 Kd (ml/g) pCi/L pCi pCi Conc pCi/g per mCi H-3 5.63E-02 0 2.25E-03 14679 0 0.00E+00 0.000E+00 C-14 1.22E-04 1.2 2.74E-04 1790.2 12888.8 3.29E-07 2.241E-02 Fe-55 2.53E-01 2857 1.31E-07 0.9 14678.1 3.74E-07 2.548E-02 Ni-59 6.86E-06 62 6.02E-06 39.4 14639.6 3.73E-07 2.541E-02 Co-60 1.31E-01 223 1.68E-06 11 14668 3.74E-07 2.548E-02 Ni-63 6.92E-03 62 6.02E-06 39.4 14639.6 3.73E-07 2.541E-02 Sr-90 2.41E-02 2.3 1.52E-04 991.8 13687.2 3.49E-07 2.378E-02 Nb-94 3.41E-05 45 8.29E-06 54.2 14624.8 3.73E-07 2.541E-02 Tc-99 3.28E-06 0 2.25E-03 14679 0 0.00E+00 0.000E+00 Ag-108m 1.66E-03 27 1.38E-05 90.1 14588.9 3.72E-07 2.534E-02 Sb-125 2.51E-01 17 2.18E-05 142.2 14536.8 3.71E-07 2.527E-02 Cs-134 3.36E-01 45 8.29E-06 54.2 14624.8 3.73E-07 2.541E-02 Cs-137 2.30E-02 45 8.29E-06 54.2 14624.8 3.73E-07 2.541E-02 Pm-147 2.64E-01 95 3.93E-06 25.7 14653.3 3.74E-07 2.548E-02 Eu-152 5.12E-02 95 3.93E-06 25.7 14653.3 3.74E-07 2.548E-02 Eu-154 7.88E-02 95 3.93E-06 25.7 14653.3 3.74E-07 2.548E-02 Eu-155 1.46E-01 95 3.93E-06 25.7 14653.3 3.74E-07 2.548E-02 Np-237 3.23E-07 1 3.21E-04 2097.1 12581.9 3.21E-07 2.187E-02 Pu-238 7.90E-03 174 2.15E-06 14 14665 3.74E-07 2.548E-02 Pu-239 2.87E-05 174 2.15E-06 14 14665 3.74E-07 2.548E-02 Pu-240 1.06E-04 174 2.15E-06 14 14665 3.74E-07 2.548E-02 Pu-241 4.83E-02 174 2.15E-06 14 14665 3.74E-07 2.548E-02 Am-241 1.60E-03 177 2.11E-06 13.8 14665.2 3.74E-07 2.548E-02 Am-243 9.40E-05 177 2.11E-06 13.8 14665.2 3.74E-07 2.548E-02 Cm-243 2.43E-02 891 4.19E-07 2.7 14676.3 3.74E-07 2.548E-02 Cm-244 3.83E-02 891 4.19E-07 2.7 14676.3 3.74E-07 2.548E-02 Page 41 of 56

Attachment A TSD 14-021 DUST MS Full Radionuclide Suite Results Revision 0 Table 41 - DUST MS Results for Crib House/Forebay Building Instantaneous Release Model with 6940 pCi Source Terms per Nuclide Peak Peak Decay Activity in Activity Peak Sorbed Constant Peak Conc Solution Sorbed Sorbed Conc pCi/g Nuclide yr-1 Kd (ml/g) pCi/L pCi pCi Conc pCi/g per mCi H-3 5.63E-02 0 9.08E-04 6936 0 0.00E+00 0.000E+00 C-14 1.22E-04 1.2 1.11E-04 845.8 6094.2 1.33E-07 1.916E-02 Fe-55 2.53E-01 2857 5.29E-08 0.4 6939.6 1.52E-07 2.190E-02 Ni-59 6.86E-06 62 2.44E-06 18.6 6921.4 1.51E-07 2.176E-02 Co-60 1.31E-01 223 6.78E-07 5.2 6934.8 1.51E-07 2.176E-02 Ni-63 6.92E-03 62 1.91E-06 14.6 6925.4 1.51E-07 2.176E-02 Sr-90 2.41E-02 2.3 6.14E-05 468.5 6471.5 1.41E-07 2.032E-02 Nb-94 3.41E-05 45 3.35E-06 25.6 6914.4 1.51E-07 2.176E-02 Tc-99 3.28E-06 0 9.09E-04 6936 0 0.00E+00 0.000E+00 Ag-108m 1.66E-03 27 5.58E-06 42.5 6897.5 1.51E-07 2.176E-02 Sb-125 2.51E-01 17 8.80E-06 67.2 6872.8 1.50E-07 2.161E-02 Cs-134 3.36E-01 45 3.35E-06 25.6 6914.4 1.51E-07 2.176E-02 Cs-137 2.30E-02 45 3.35E-06 25.6 6914.4 1.51E-07 2.176E-02 Pm-147 2.64E-01 95 1.59E-06 12.1 6927.9 1.51E-07 2.176E-02 Eu-152 5.12E-02 95 1.59E-06 12.1 6927.9 1.51E-07 2.176E-02 Eu-154 7.88E-02 95 1.59E-06 12.1 6927.9 1.51E-07 2.176E-02 Eu-155 1.46E-01 95 1.59E-06 12.1 6927.9 1.51E-07 2.176E-02 Np-237 3.23E-07 1 1.30E-04 990.8 5949.2 1.30E-07 1.873E-02 Pu-238 7.90E-03 174 8.70E-07 6.6 6933.4 1.51E-07 2.176E-02 Pu-239 2.87E-05 174 8.70E-07 6.6 6933.4 1.51E-07 2.176E-02 Pu-240 1.06E-04 174 8.70E-07 6.6 6933.4 1.51E-07 2.176E-02 Pu-241 4.83E-02 174 8.70E-07 6.6 6933.4 1.51E-07 2.176E-02 Am-241 1.60E-03 177 8.55E-07 6.5 6933.5 1.51E-07 2.176E-02 Am-243 9.40E-05 177 8.55E-07 6.5 6933.5 1.51E-07 2.176E-02 Cm-243 2.43E-02 891 1.70E-07 1.3 6938.7 1.52E-07 2.190E-02 Cm-244 3.83E-02 891 1.70E-07 1.3 6938.7 1.52E-07 2.190E-02 Page 42 of 56

Attachment A TSD 14-021 DUST MS Full Radionuclide Suite Results Revision 0 Table 42 - DUST MS Results for Waste Water Treatment Facility Instantaneous Release Model with 1124 pCi Source Terms per Nuclide Peak Peak Decay Activity in Activity Peak Sorbed Constant Peak Conc Solution Sorbed Sorbed Conc pCi/g Nuclide yr-1 Kd (ml/g) pCi/L pCi pCi Conc pCi/g per mCi H-3 5.63E-02 0 3.13E-02 1126 0 0.00E+00 0.000E+00 C-14 1.22E-04 1.2 3.82E-03 137.5 990.3 4.58E-06 4.075E+00 Fe-55 2.53E-01 2857 1.82E-06 0.1 1124.9 5.21E-06 4.635E+00 Ni-59 6.86E-06 62 8.40E-05 3 1124.8 5.21E-06 4.635E+00 Co-60 1.31E-01 223 2.34E-05 0.8 1125.5 5.21E-06 4.635E+00 Ni-63 6.92E-03 62 8.40E-05 3 1124.8 5.21E-06 4.635E+00 Sr-90 2.41E-02 2.3 2.12E-03 76.2 1051.4 4.87E-06 4.333E+00 Nb-94 3.41E-05 45 1.16E-04 4.2 1123.7 5.20E-06 4.626E+00 Tc-99 3.28E-06 0 3.13E-02 1128 0 0.00E+00 0.000E+00 Ag-108m 1.66E-03 27 1.92E-04 6.9 1120.9 5.19E-06 4.617E+00 Sb-125 2.51E-01 17 3.03E-04 10.9 1114.1 5.16E-06 4.591E+00 Cs-134 3.36E-01 45 1.16E-04 4.2 1123.4 5.20E-06 4.626E+00 Cs-137 2.30E-02 45 1.16E-04 4.2 1123.4 5.20E-06 4.626E+00 Pm-147 2.64E-01 95 5.48E-05 2 1125.3 5.21E-06 4.635E+00 Eu-152 5.12E-02 95 5.48E-05 2 1125.3 5.21E-06 4.635E+00 Eu-154 7.88E-02 95 5.48E-05 2 1125.3 5.21E-06 4.635E+00 Eu-155 1.46E-01 95 5.48E-05 2 1125.3 5.21E-06 4.635E+00 Np-237 3.23E-07 1 4.48E-03 161.1 966.7 4.48E-06 3.986E+00 Pu-238 7.90E-03 174 3.00E-05 1.1 1126.8 5.22E-06 4.644E+00 Pu-239 2.87E-05 174 3.00E-05 1.1 1126.8 5.22E-06 4.644E+00 Pu-240 1.06E-04 174 3.00E-05 1.1 1126.8 5.22E-06 4.644E+00 Pu-241 4.83E-02 174 3.00E-05 1.1 1126.8 5.22E-06 4.644E+00 Am-241 1.60E-03 177 2.95E-05 1.1 1126.8 5.22E-06 4.644E+00 Am-243 9.40E-05 177 2.95E-05 1.1 1126.8 5.22E-06 4.644E+00 Cm-243 2.43E-02 891 5.85E-06 0.2 1124.9 5.21E-06 4.635E+00 Cm-244 3.83E-02 891 5.85E-06 0.2 1124.9 5.21E-06 4.635E+00 Page 43 of 56

Attachment B TSD 14-021 Construction Scenario MicroShield Calculations Revision 0 MicroShield 8.03 Radiation Safety and Control Services (8.03-0000)

Date By Checked Filename Run Date Run Time Duration AUXINTRUDER.msd December 11, 2014 12:56:21 PM 00:00:00 Project Info Case Title Zion Intruder Description Aux Building Construction Foundation Excavation Geometry 13 - Rectangular Volume Source Dimensions Length 1.1e+3 cm (37 ft 0.9 in)

Width 8.0e+3 cm (263 ft)

Height 3.1e+3 cm (103 ft 0.0 in)

Dose Points A X Y Z 1.2e+3 cm (39 ft 3.3 1.6e+3 cm (51 ft 6.0 4.0e+3 cm (131 ft 6.0

1 in) in) in) 1.3e+3 cm (41 ft 2.8 1.6e+3 cm (51 ft 6.0 4.0e+3 cm (131 ft 6.0

2 in) in) in)

Shields Shield N Dimension Material Density Source 1.00e+06 ft³ Concrete 1.5 Shield 1 2.16 ft Concrete 1.5 Air Gap Air 0.00122 Source Input: Grouping Method - Actual Photon Energies Nuclide Ci Bq µCi/cm³ Bq/cm³ Ba-137m 5.7148e-001 2.1145e+010 2.0095e-005 7.4353e-001 Co-60 3.8310e-004 1.4175e+007 1.3471e-008 4.9844e-004 Cs-137 6.0410e-001 2.2352e+010 2.1243e-005 7.8598e-001 Buildup: The material reference is Shield 1 Page 44 of 56

Attachment B TSD 14-021 Construction Scenario MicroShield Calculations Revision 0 Integration Parameters X Direction 10 Y Direction 20 Z Direction 20 Results - Dose Point # 1 - (3.93e+01,51.5,131.5) ft Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec) MeV/cm²/sec MeV/cm²/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.0045 2.195e+08 0.000e+00 3.026e-28 0.000e+00 2.074e-28 0.0318 4.378e+08 1.441e-65 8.441e-27 1.200e-67 7.031e-29 0.0322 8.077e+08 1.865e-63 1.618e-26 1.501e-65 1.303e-28 0.0364 2.939e+08 4.320e-48 9.019e-27 2.454e-50 5.124e-29 0.6616 1.903e+10 4.572e-05 1.657e-03 8.864e-08 3.213e-06 0.6938 2.312e+03 7.412e-12 2.490e-10 1.431e-14 4.808e-13 1.1732 1.417e+07 9.246e-07 1.397e-05 1.652e-09 2.496e-08 1.3325 1.417e+07 1.819e-06 2.304e-05 3.156e-09 3.998e-08 Totals 2.081e+10 4.847e-05 1.694e-03 9.345e-08 3.278e-06 Results - Dose Point # 2 - (4.12e+01,51.5,131.5) ft Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec) MeV/cm²/sec MeV/cm²/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.0045 2.195e+08 0.000e+00 2.781e-28 0.000e+00 1.906e-28 0.0318 4.378e+08 6.667e-59 7.758e-27 5.554e-61 6.462e-29 0.0322 8.077e+08 5.351e-57 1.488e-26 4.307e-59 1.197e-28 0.0364 2.939e+08 2.078e-43 8.290e-27 1.181e-45 4.710e-29 0.6616 1.903e+10 8.152e-05 2.561e-03 1.580e-07 4.964e-06 0.6938 2.312e+03 1.294e-11 3.786e-10 2.499e-14 7.310e-13 1.1732 1.417e+07 1.319e-06 1.797e-05 2.358e-09 3.211e-08 1.3325 1.417e+07 2.489e-06 2.862e-05 4.318e-09 4.966e-08 Totals 2.081e+10 8.533e-05 2.607e-03 1.647e-07 5.046e-06 MicroShield 8.03 Radiation Safety and Control Services (8.03-0000)

Page 45 of 56

Attachment B TSD 14-021 Construction Scenario MicroShield Calculations Revision 0 Date By Checked Filename Run Date Run Time Duration SFBINTRUDERSLAB.msd December 11, 2014 1:14:33 PM 00:00:01 Project Info Case Title Zion Intruder Description Spent Fuel Building Construction Foundation Excavation Slab Geometry 13 - Rectangular Volume Source Dimensions Length 91.44 cm (3 ft)

Width 1.9e+3 cm (63 ft)

Height 1.2e+3 cm (39 ft 0.0 in)

Dose Points A X Y Z 168.402 cm (5 ft 6.3 594.36 cm (19 ft 6.0 960.12 cm (31 ft 6.0

1 in) in) in) 228.092 cm (7 ft 5.8 594.36 cm (19 ft 6.0 960.12 cm (31 ft 6.0

2 in) in) in)

Shields Shield N Dimension Material Density Source 7371.0 ft³ Concrete 1.5 Shield 1 2.15 ft Concrete 1.5 Shield 2 .333 ft Concrete 2.35 Air Gap Air 0.00122 Source Input: Grouping Method - Actual Photon Energies Nuclide Ci Bq µCi/cm³ Bq/cm³ Ba-137m 3.4520e-001 1.2772e+010 1.6538e-003 6.1192e+001 Co-60 2.3260e-004 8.6062e+006 1.1144e-006 4.1233e-002 Cs-137 3.6490e-001 1.3501e+010 1.7482e-003 6.4685e+001 Buildup: The material reference is Shield 2 Integration Parameters Page 46 of 56

Attachment B TSD 14-021 Construction Scenario MicroShield Calculations Revision 0 X Direction 10 Y Direction 20 Z Direction 20 Results - Dose Point # 1 - (5.525,19.5,31.5) ft Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec) MeV/cm²/sec MeV/cm²/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.0045 1.326e+08 0.000e+00 2.434e-27 0.000e+00 1.668e-27 0.0318 2.644e+08 6.883e-57 6.789e-26 5.734e-59 5.655e-28 0.0322 4.879e+08 4.936e-55 1.302e-25 3.973e-57 1.048e-27 0.0364 1.775e+08 7.368e-42 7.254e-26 4.186e-44 4.121e-28 0.6616 1.149e+10 1.089e-03 3.908e-02 2.111e-06 7.575e-05 0.6938 1.404e+03 1.770e-10 5.921e-09 3.417e-13 1.143e-11 1.1732 8.606e+06 2.262e-05 3.564e-04 4.042e-08 6.369e-07 1.3325 8.606e+06 4.534e-05 6.038e-04 7.865e-08 1.048e-06 Totals 1.257e+10 1.157e-03 4.004e-02 2.230e-06 7.744e-05 Results - Dose Point # 2 - (7.48e+00,19.5,31.5) ft Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec) MeV/cm²/sec MeV/cm²/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.0045 1.326e+08 0.000e+00 1.935e-27 0.000e+00 1.326e-27 0.0318 2.644e+08 9.599e-57 5.397e-26 7.996e-59 4.495e-28 0.0322 4.879e+08 6.615e-55 1.035e-25 5.324e-57 8.327e-28 0.0364 1.775e+08 7.176e-42 5.766e-26 4.077e-44 3.276e-28 0.6616 1.149e+10 1.096e-03 3.968e-02 2.124e-06 7.693e-05 0.6938 1.404e+03 1.783e-10 6.016e-09 3.443e-13 1.162e-11 1.1732 8.606e+06 2.299e-05 3.623e-04 4.107e-08 6.474e-07 1.3325 8.606e+06 4.607e-05 6.130e-04 7.994e-08 1.064e-06 Totals 1.257e+10 1.165e-03 4.066e-02 2.245e-06 7.864e-05 Page 47 of 56

Attachment B TSD 14-021 Open Air Demo Cut-Off MicroShield Calculations Revision 0 MicroShield 8.03 Radiation Safety and Control Services (8.03-0000)

Date By Checked Filename Run Date Run Time Duration AUX Core 1.msd June 30, 2014 4:19:01 AM 00:00:00 Project Info Case Title Aux Worst Description First 0 - 0.5 in Eight Inch Diameter Geometry 8 - Cylinder Volume - End Shields Source Dimensions Height 1.27 cm (0.5 in)

Radius 10.16 cm (4.0 in)

Dose Points A X Y Z

1 0.0 cm (0 in) 2.54 cm (1.0 in) 0.0 cm (0 in)

Shields Shield N Dimension Material Density Source 25.133 in³ Concrete 2.35 Air Gap Air 0.00122 Source Input: Grouping Method - Actual Photon Energies Nuclide Ci Bq µCi/cm³ Bq/cm³ Ba-137m 1.8447e-005 6.8254e+005 4.4790e-002 1.6572e+003 Co-60 1.9000e-007 7.0300e+003 4.6133e-004 1.7069e+001 Cs-137 1.9500e-005 7.2150e+005 4.7347e-002 1.7518e+003 Buildup: The material reference is Source Integration Parameters Radial 20 Circumferential 10 Y Direction (axial) 10 Results Energy (MeV) Activity (Photons/sec) Fluence Rate Fluence Rate Exposure Rate Exposure Rate Page 48 of 56

Attachment B TSD 14-021 Open Air Demo Cut-Off MicroShield Calculations Revision 0 MeV/cm²/sec MeV/cm²/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.0045 7.085e+03 1.765e-03 1.807e-03 1.210e-03 1.239e-03 0.0318 1.413e+04 2.039e-01 2.458e-01 1.698e-03 2.047e-03 0.0322 2.607e+04 3.915e-01 4.746e-01 3.151e-03 3.820e-03 0.0364 9.487e+03 2.119e-01 2.732e-01 1.204e-03 1.552e-03 0.6616 6.141e+05 8.334e+02 1.021e+03 1.616e+00 1.979e+00 0.6938 1.147e+00 1.640e-03 1.995e-03 3.166e-06 3.852e-06 1.1732 7.030e+03 1.791e+01 2.050e+01 3.200e-02 3.663e-02 1.3325 7.030e+03 2.057e+01 2.325e+01 3.569e-02 4.034e-02 Totals 6.850e+05 8.727e+02 1.065e+03 1.691e+00 2.064e+00 MicroShield 8.03 Radiation Safety and Control Services (8.03-0000)

Date By Checked Filename Run Date Run Time Duration AUX Core 2.msd June 30, 2014 4:21:14 AM 00:00:00 Project Info Case Title Aux Worst Description First 0.5-1.0 in 8 Inch Diameter Geometry 8 - Cylinder Volume - End Shields Source Dimensions Height 1.27 cm (0.5 in)

Radius 10.16 cm (4.0 in)

Dose Points A X Y Z

1 0.0 cm (0 in) 3.81 cm (1.5 in) 0.0 cm (0 in)

Shields Shield N Dimension Material Density Source 25.133 in³ Concrete 2.35 Shield 1 .5 in Concrete 2.35 Page 49 of 56

Attachment B TSD 14-021 Open Air Demo Cut-Off MicroShield Calculations Revision 0 Air Gap Air 0.00122 Source Input: Grouping Method - Actual Photon Energies Nuclide Ci Bq µCi/cm³ Bq/cm³ Ba-137m 1.0406e-005 3.8502e+005 2.5266e-002 9.3486e+002 Co-60 1.6700e-007 6.1790e+003 4.0549e-004 1.5003e+001 Cs-137 1.1000e-005 4.0700e+005 2.6709e-002 9.8822e+002 Buildup: The material reference is Shield 1 Integration Parameters Radial 20 Circumferential 10 Y Direction (axial) 10 Results Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec) MeV/cm²/sec MeV/cm²/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.0045 3.997e+03 1.497e-15 1.649e-15 1.026e-15 1.130e-15 0.0318 7.971e+03 1.501e-03 2.302e-03 1.250e-05 1.917e-05 0.0322 1.471e+04 3.201e-03 4.953e-03 2.576e-05 3.986e-05 0.0364 5.352e+03 4.323e-03 7.386e-03 2.456e-05 4.196e-05 0.6616 3.464e+05 2.281e+02 3.737e+02 4.422e-01 7.244e-01 0.6938 1.008e+00 7.048e-04 1.137e-03 1.361e-06 2.195e-06 1.1732 6.179e+03 8.346e+00 1.165e+01 1.491e-02 2.082e-02 1.3325 6.179e+03 9.763e+00 1.321e+01 1.694e-02 2.292e-02 Totals 3.908e+05 2.462e+02 3.986e+02 4.741e-01 7.683e-01 MicroShield 8.03 Radiation Safety and Control Services (8.03-0000)

Date By Checked Filename Run Date Run Time Duration AUX Core 3.msd June 30, 2014 4:23:22 AM 00:00:00 Project Info Case Title Aux Worst Page 50 of 56

Attachment B TSD 14-021 Open Air Demo Cut-Off MicroShield Calculations Revision 0 Description First 1.0 - 1.5 in 8 Inch Diameter Geometry 8 - Cylinder Volume - End Shields Source Dimensions Height 1.27 cm (0.5 in)

Radius 10.16 cm (4.0 in)

Dose Points A X Y Z

1 0.0 cm (0 in) 5.08 cm (2.0 in) 0.0 cm (0 in)

Shields Shield N Dimension Material Density Source 25.133 in³ Concrete 2.35 Shield 1 1.0 in Concrete 2.35 Air Gap Air 0.00122 Source Input: Grouping Method - Actual Photon Energies Nuclide Ci Bq µCi/cm³ Bq/cm³ Ba-137m 4.9665e-006 1.8376e+005 1.2059e-002 4.4618e+002 Co-60 1.2100e-007 4.4770e+003 2.9379e-004 1.0870e+001 Cs-137 5.2500e-006 1.9425e+005 1.2747e-002 4.7165e+002 Buildup: The material reference is Shield 1 Integration Parameters Radial 20 Circumferential 10 Y Direction (axial) 10 Results Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec) MeV/cm²/sec MeV/cm²/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.0045 1.908e+03 1.266e-26 1.417e-26 8.680e-27 9.709e-27 0.0318 3.804e+03 2.292e-05 3.827e-05 1.909e-07 3.187e-07 0.0322 7.019e+03 5.344e-05 9.019e-05 4.301e-07 7.259e-07 0.0364 2.554e+03 1.553e-04 2.960e-04 8.825e-07 1.682e-06 0.6616 1.653e+05 6.257e+01 1.236e+02 1.213e-01 2.397e-01 Page 51 of 56

Attachment B TSD 14-021 Open Air Demo Cut-Off MicroShield Calculations Revision 0 0.6938 7.303e-01 2.952e-04 5.716e-04 5.699e-07 1.104e-06 1.1732 4.477e+03 3.706e+00 5.908e+00 6.624e-03 1.056e-02 1.3325 4.477e+03 4.392e+00 6.716e+00 7.620e-03 1.165e-02 Totals 1.896e+05 7.067e+01 1.363e+02 1.356e-01 2.619e-01 MicroShield 8.03 Radiation Safety and Control Services (8.03-0000)

Date By Checked Filename Run Date Run Time Duration AUX Core 4.msd June 30, 2014 4:25:26 AM 00:00:00 Project Info Case Title Aux Worst Description First 1.0 - 1.5 in 8 Inch Diameter Geometry 8 - Cylinder Volume - End Shields Source Dimensions Height 1.27 cm (0.5 in)

Radius 10.16 cm (4.0 in)

Dose Points A X Y Z

1 0.0 cm (0 in) 5.08 cm (2.0 in) 0.0 cm (0 in)

Shields Shield N Dimension Material Density Source 25.133 in³ Concrete 2.35 Shield 1 1.0 in Concrete 2.35 Air Gap Air 0.00122 Source Input: Grouping Method - Actual Photon Energies Nuclide Ci Bq µCi/cm³ Bq/cm³ Ba-137m 2.0244e-006 7.4904e+004 4.9155e-003 1.8187e+002 Co-60 7.6400e-008 2.8268e+003 1.8550e-004 6.8636e+000 Cs-137 2.1400e-006 7.9180e+004 5.1960e-003 1.9225e+002 Page 52 of 56

Attachment B TSD 14-021 Open Air Demo Cut-Off MicroShield Calculations Revision 0 Buildup: The material reference is Shield 1 Integration Parameters Radial 20 Circumferential 10 Y Direction (axial) 10 Results Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec) MeV/cm²/sec MeV/cm²/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.0045 7.776e+02 5.162e-27 5.774e-27 3.538e-27 3.958e-27 0.0318 1.551e+03 9.343e-06 1.560e-05 7.783e-08 1.299e-07 0.0322 2.861e+03 2.178e-05 3.676e-05 1.753e-07 2.959e-07 0.0364 1.041e+03 6.332e-05 1.206e-04 3.597e-07 6.854e-07 0.6616 6.740e+04 2.551e+01 5.040e+01 4.945e-02 9.770e-02 0.6938 4.611e-01 1.864e-04 3.609e-04 3.599e-07 6.968e-07 1.1732 2.827e+03 2.340e+00 3.731e+00 4.182e-03 6.667e-03 1.3325 2.827e+03 2.773e+00 4.241e+00 4.811e-03 7.357e-03 Totals 7.928e+04 3.062e+01 5.837e+01 5.844e-02 1.117e-01 MicroShield 8.03 Radiation Safety and Control Services (8.03-0000)

Date By Checked Filename Run Date Run Time Duration AUX Core 5.msd June 30, 2014 4:27:22 AM 00:00:00 Project Info Case Title Aux Worst Description First 1.5 - 2.0 in 8 Inch Diameter Geometry 8 - Cylinder Volume - End Shields Page 53 of 56

Attachment B TSD 14-021 Open Air Demo Cut-Off MicroShield Calculations Revision 0 Source Dimensions Height 1.27 cm (0.5 in)

Radius 10.16 cm (4.0 in)

Dose Points A X Y Z

1 0.0 cm (0 in) 6.35 cm (2.5 in) 0.0 cm (0 in)

Shields Shield N Dimension Material Density Source 25.133 in³ Concrete 2.35 Shield 1 1.5 in Concrete 2.35 Air Gap Air 0.00122 Source Input: Grouping Method - Actual Photon Energies Nuclide Ci Bq µCi/cm³ Bq/cm³ Ba-137m 1.6271e-006 6.0203e+004 3.9507e-003 1.4618e+002 Co-60 8.0400e-008 2.9748e+003 1.9522e-004 7.2230e+000 Cs-137 1.7200e-006 6.3640e+004 4.1763e-003 1.5452e+002 Buildup: The material reference is Shield 1 Integration Parameters Radial 20 Circumferential 10 Y Direction (axial) 10 Results Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec) MeV/cm²/sec MeV/cm²/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.0045 6.250e+02 9.574e-38 2.985e-29 6.563e-38 2.046e-29 0.0318 1.246e+03 2.879e-07 5.112e-07 2.398e-09 4.258e-09 0.0322 2.300e+03 7.327e-07 1.316e-06 5.896e-09 1.059e-08 0.0364 8.368e+02 4.516e-06 9.271e-06 2.566e-08 5.267e-08 0.6616 5.417e+04 1.259e+01 2.898e+01 2.441e-02 5.619e-02 0.6938 4.852e-01 1.211e-04 2.721e-04 2.338e-07 5.253e-07 1.1732 2.975e+03 1.602e+00 2.847e+00 2.862e-03 5.087e-03 Page 54 of 56

Attachment B TSD 14-021 Open Air Demo Cut-Off MicroShield Calculations Revision 0 1.3325 2.975e+03 1.920e+00 3.247e+00 3.330e-03 5.634e-03 Totals 6.513e+04 1.611e+01 3.508e+01 3.060e-02 6.691e-02 MicroShield 8.03 Radiation Safety and Control Services (8.03-0000)

Date By Checked Filename Run Date Run Time Duration AUX Core 6.msd June 30, 2014 4:29:47 AM 00:00:00 Project Info Case Title Aux Worst Description First 2.0 - 2.5 in 8 Inch Diameter Geometry 8 - Cylinder Volume - End Shields Source Dimensions Height 1.27 cm (0.5 in)

Radius 10.16 cm (4.0 in)

Dose Points A X Y Z

1 0.0 cm (0 in) 7.62 cm (3.0 in) 0.0 cm (0 in)

Shields Shield N Dimension Material Density Source 25.133 in³ Concrete 2.35 Shield 1 2.0 in Concrete 2.35 Air Gap Air 0.00122 Source Input: Grouping Method - Actual Photon Energies Nuclide Ci Bq µCi/cm³ Bq/cm³ Ba-137m 1.0879e-006 4.0252e+004 2.6415e-003 9.7735e+001 Co-60 4.6200e-008 1.7094e+003 1.1218e-004 4.1505e+000 Cs-137 1.1500e-006 4.2550e+004 2.7923e-003 1.0331e+002 Buildup: The material reference is Shield 1 Integration Parameters Radial 20 Page 55 of 56

Attachment B TSD 14-021 Open Air Demo Cut-Off MicroShield Calculations Revision 0 Circumferential 10 Y Direction (axial) 10 Results Fluence Rate Fluence Rate Exposure Rate Exposure Rate Energy (MeV) Activity (Photons/sec) MeV/cm²/sec MeV/cm²/sec mR/hr mR/hr No Buildup With Buildup No Buildup With Buildup 0.0045 4.179e+02 1.651e-48 1.592e-29 1.132e-48 1.091e-29 0.0318 8.333e+02 8.033e-09 1.495e-08 6.691e-11 1.245e-10 0.0322 1.538e+03 2.230e-08 4.205e-08 1.795e-10 3.384e-10 0.0364 5.595e+02 2.899e-07 6.319e-07 1.647e-09 3.590e-09 0.6616 3.622e+04 5.366e+00 1.412e+01 1.040e-02 2.738e-02 0.6938 2.788e-01 4.457e-05 1.141e-04 8.604e-08 2.204e-07 1.1732 1.709e+03 6.192e-01 1.213e+00 1.107e-03 2.168e-03 1.3325 1.709e+03 7.502e-01 1.389e+00 1.302e-03 2.411e-03 Totals 4.299e+04 6.736e+00 1.672e+01 1.281e-02 3.196e-02 Page 56 of 56

Technical Support Document TSD 14-021 BFM Drilling Spoils and Alternate Exposure Scenarios Revision 0 Originator: ___ ______________________________ Date: __12/18/14_______

Harvey Farr Reviewer: __(signature on file)___________________ Date:__12/18/14_______

Dave Fauver Approval: __(signature on file)___________________ Date:__12/18/14_______

Robert F. Yetter

TSD 14-021 Revision 0 Summary of Changes in this Revision:

Rev. 0 -Initial issuance.