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TSD 14-009 Revision 1 Page 1 of 56
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TSD 14-009 Revision 1 BNL-107250-2014-IR_R1 EVALUATION OF MAXIMUM RADIONUCLIDE GROUNDWATER CONCENTRATIONS FOR BASEMENT FILL MODEL Zion Station Restoration Project Terry Sullivan December 2, 2014 Informal Report Biological, Environmental & Climate Sciences Department Brookhaven National Laboratory P.O. Box 5000 Upton, NY 11973-5000 www.bnl.gov Notice: This manuscript has been authored by employees of Brookhaven Science Associates, LLC under Contract No. DE -AC02-98CH10886 with the U.S. Department of Energy. The publisher by accepting the manuscript for publication acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.
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TSD 14-009 Revision 1 DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, nor any of their contractors, subcontractors, or their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or any third partys use or the results of such use of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights.
Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof or its contractors or subcontractors. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.
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TSD 14-009 Revision 1 Table of Contents 1.Introduction .................................................................................................................................... 1 2.Conceptual Models of Release ....................................................................................................... 3 2.1 Site Overview.......................................................................................................................... 3 2.2 Modeling Overview ................................................................................................................ 3 2.3 Release Models ....................................................................................................................... 9 2.3.1 Instant Release ................................................................................................................. 9 2.3.2 Release Rate: Diffusion Controlled Release from the concrete...................................... 9 2.4 Receptor Well Outside the Turbine Building........................................................................ 10 3.Analysis Parameters ..................................................................................................................... 12 3.1 Parameters ............................................................................................................................. 12 3.1.1 Diffusion Controlled Release Model ............................................................................. 13 3.1.2 Model Geometry ............................................................................................................ 13 3.1.3 Receptor Well Parameters for Transport Model ............................................................ 14 3.1.4 Sensitivity Analysis Parameters..................................................................................... 15 4 Results .......................................................................................................................................... 17 4.1 Base Case Release Peak Groundwater Concentration Results ............................................. 17 4.1.1 Auxiliary Building ......................................................................................................... 17 4.1.2 Containment Buildings .................................................................................................. 18 4.1.3 Crib House/Forebay ....................................................................................................... 18 4.1.4 Fuel Building ................................................................................................................. 19 4.1.5 Turbine Building ............................................................................................................ 20 4.1.6 Waste Water Treatment Facility .................................................................................... 20 4.2 Sensitivity Analysis .............................................................................................................. 21 4.2.1 Sensitivity to Release Rate ............................................................................................ 22 4.2.2 Drill Spoils Sensitivity to Kd ......................................................................................... 24 4.3 Outside Receptor Well Concentration in Transport Model .................................................. 25 4.4 Discussion ............................................................................................................................. 26 5 Validation..................................................................................................................................... 27 6 Conclusions .................................................................................................................................. 28 7 References .................................................................................................................................... 29 Appendix A: Sensitivity Analysis Results ..................................................................................... 31 Page 5 of 56
TSD 14-009 Revision 1 A.1: Base Case ........................................................................................................................... 31 A.2: High Kd.............................................................................................................................. 32 A.3: Low Kd .............................................................................................................................. 33 A.4: High Porosity ..................................................................................................................... 34 A.5: Low Porosity ...................................................................................................................... 35 A.6: High Bacfkill Density ........................................................................................................ 36 A.7: Low Density ....................................................................................................................... 37 Appendix B: Calculations to address Request for Additional Information (RAI) - 21. ................ 38 B1. Introduction .............................................................................................................................. 38 B2. Conceptual Models of Release ................................................................................................ 39 B2.1 Site Overview ..................................................................................................................... 39 B2.2 Modeling Overview ........................................................................................................... 40 B2.3 Release Models .................................................................................................................. 40 B2.4 Model Geometry ............................................................................................................... 40 B-2.4.1 Receptor Well ................................................................................................................ 42 B3. Analysis Parameters ................................................................................................................. 43 B3.1 Parameters .......................................................................................................................... 43 B-3.1.1 Source Term Inventory .............................................................................................. 44 B.3.1.2 Receptor Well Parameters for Transport Model ........................................................ 46 B4 Results ....................................................................................................................................... 47 B4.1 Base Case .......................................................................................................................... 47 B5 Conclusions ............................................................................................................................... 48 B6 References ............................................................................................................................. 49 Figures Figure 1 Zion Site building layout .................................................................................................. 5 Figure 2. Geometry of the Auxiliary Building. ............................................................................. 8 Figure 3 Schematic Representation of Flow the geometry used to assess flow to a well outside the Turbine Building. ................................................................................................................... 11 Page 6 of 56
TSD 14-009 Revision 1 Tables Table 1 Mixing volume and release rate assumption..................................................................... 6 Table 2 Geometric Parameters and Unit Inventory for Residual Structures (Farr, 2014) ............. 8 Table 3 Potential Radionuclides of Concern at the Zion Nuclear Power Station .......................... 9 Table 4 Typical diffusion coefficients in cement for radionuclides of concern ............................ 9 Table 5 Selected distribution coefficients (Sullivan, 2014) ......................................................... 13 Table 6 Model Geometry for all simulations ............................................................................... 14 Table 7 Transport Parameters used to calculate peak concentrations in a receptor well located outside of the basements ............................................................................................................... 14 Table 8 Parameters and their range in the sensitivity analysis .................................................... 15 Table 9 Kd values selected to examine the sensitivity of drill spoils predicted soil and groundwater concentrations .......................................................................................................... 16 Table 10 Auxiliary Building Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2 and diffusion controlled release from 0.5 inch of contaminated concrete. The total inventory for each radionuclide is 6503 pCi ................................................................................. 17 Table 11 Containment Building Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2. The total inventory for each radionuclide is 2759 pCi ................................................... 18 Table 12 Crib House Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2.
The total inventory for each radionuclide is 6940 pCi ................................................................. 19 Table 13 Fuel Building Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2.
Release is diffusion controlled from 0.5 inch thick contaminated region. The total inventory for each radionuclide is 780 pCi ......................................................................................................... 19 Table 14 Turbine Building Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2. The total inventory for each radionuclide is 14679 pCi ................................................. 20 Table 15 Waste Water Treatment Facility Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2. The total inventory for each radionuclide is 1124 pCi ................................ 21 Table 16 Comparison of the percentage of the total inventory released based on the thickness of the contaminated zone. Thicknesses analyzed were 0.5 inch (base case), 1 inch and 2 inch. ..... 22 Table 17 Comparison of the peak water concentration based on the thickness of the contaminated zone. Thicknesses analyzed were 1 inch (base case), 1/2 and 2 inch. ........................................... 23 Table 18 Comparison of the time to reach the peak concentration in solution based on the thickness of the contaminated zone. Thicknesses analyzed were 0.5 inch (base case), 1 inch and 2 inch............................................................................................................................................. 24 Table 19 Sensitivity of Drill Spoils to Distribution Coefficient (Kd) .......................................... 24 Table 20 Comparison of Base Case and Drill Spoils case ........................................................... 25 Table 20 Comparison of Peak Concentrations in the modeled region......................................... 26 Table 21 Comparison between Analytical Solution and DUST-MS results for the Turbine Building. ....................................................................................................................................... 27 Page 7 of 56
TSD 14-009 Revision 1 Evaluation of Maximum Radionuclide Groundwater Concentrations for Radionuclides of Concern Zion Station Restoration Project
- 1. Introduction ZionSolutions is in the process of decommissioning the Zion Nuclear Power Station (ZNPS).
After decommissioning is completed, the site will contain two reactor Containment Buildings, the Fuel Handling Building and Transfer Canals, Auxiliary Building, Turbine Building, Crib House/Forebay, and a Waste Water Treatment Facility that have been demolished to a depth of 3 feet below grade. Additional below ground structures remaining will include the Main Steam Tunnels and large diameter intake and discharge pipes. These additional structures are not included in the modeling described in this report but the inventory remaining (expected to be very low) will be included with one of the structures that are modeled as designated in the Zion Station Restoration Project (ZSRP) License Termination Plan (LTP). The remaining underground structures will be backfilled with clean material. The final selection of fill material has not been made.
Remaining structures will contain residual radioactive material to varying extents. The bulk of the source term will be contained in the concrete floors. Current interior demolition plans are to remove all concrete inside the steel liner in the Unit 1 and Unit 2 Containment Buildings. Based upon concrete characterization data, the highest end state source term is anticipated to be contained in the Auxiliary Building floor located approximately 50 feet below grade. The end state source term will be at least 3 feet below grade in all remaining structures eliminating conventional pathways such as direct radiation and inhalation rendering groundwater related pathways the most significant potential sources of future exposure.
An important component of the decommissioning process is the demonstration that any remaining activity will not cause a hypothetical individual (average member of the critical group) to receive a dose in excess of 25 mrem/y as specified in 10 CFR Part 20 Subpart E. To demonstrate compliance with 10 CFR Part 20 Subpart E requires modeling of the fate and transport of radioactive material to a receptor. This involves characterization of the building basements to remain on site to quantify the amount of residual radioactivity, modeling the release of radioactivity from the concrete, and mixing with the water contained in the fill material. Transport away from the fill to a receptor well located outside of the basements may also be a relevant pathway.
A previous study (Sullivan, 2014a) performed screening calculations for the Auxiliary Building for 26 radionuclides. The Auxiliary Building was used for the screening calculations because it is expected to contain the majority of the residual contamination inventory at the time of license termination. This analysis was used by ZSRP along with characterization data and RESRAD modeling to screen out low dose significance radionuclides and identify eight radionuclides of concern (ROCs) Co-60, Ni-63, Sr-90, Cs-134, Cs-137, Eu-152, Eu-154, and H-3 for detailed assessment.
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TSD 14-009 Revision 1 This report addresses the release of a given radionuclide inventory, for each of the ROCs, to the interstitial water of the fill material and calculates the equilibrium concentration at a well located in the middle of the subsurface remains of the seven buildings. The ratio of the resulting equilibrium water concentration in units of picocuries per liter (pCi/L) to the assumed inventory in units of Curies (Ci) for each building is used by ZSRP, in conjunction with the RESRAD code, to demonstrate compliance with 10 CFR 20 Subpart E.
Calculation of the fill interstitial water concentration requires site-specific information on the hydrogeologic properties (effective porosity and bulk density) and chemical transport properties (sorption). Conestoga-Rovers & Associates (CRA) has collected a substantial amount of site-specific hydrogeologic data (CRA, 2014).
Brookhaven National Laboratory (BNL) has determined site-specific sorption data for five nuclides that are ROCs with four soil types, two concrete types of construction demolition debris, two cinder block materials, and one grout material that are under consideration for the fill (Yim, 2012, Milian, 2014). Two ROCs, Eu-152 and Eu-154 have not had site-specific sorption measurements. A report (Sullivan, 2014) provided recommended values to use for dose assessment based on measured values, when available, and literature values in other cases. For nuclides with site-specific measured values, the lowest measured distribution coefficient in any of the media tested was recommended for use.
The objectives of this report are:
a) To present a simplified conceptual model for release from the buildings with residual subsurface structures that can be used to provide an upper bound on contaminant concentrations in the fill material.
b) Provide maximum water concentrations and the corresponding amount of mass sorbed to the solid fill material that could occur in each building for use in dose assessment calculations.
c) Estimate the maximum concentration in a well located outside of the fill material.
d) Perform a sensitivity analysis of key parameters.
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TSD 14-009 Revision 1
- 2. Conceptual Models of Release 2.1 Site Overview Figure 1 provides the site layout at ZNPS located on the shores of Lake Michigan. Major features include two reactor Containment Buildings (Unit-1 and Unit-2 in Figure 1, a Fuel Handling Building, Auxiliary Building, Turbine Building, Crib House, and Waste Water Treatment Facility (WWTF).
The proposed decommissioning approach involves removal of regions with high-levels of contamination through a remediation process. There will be some surface contamination and volumetric contamination left in place. This contamination will provide a potential source of radioactivity to the groundwater. These structures will be filled with non-contaminated material.
Fills that have been under consideration include:
Clean concrete construction debris (CCDD);
Clean cinder block material; Clean Sand Clean Grout Recently, grout has been eliminated from consideration for fill material. The fill may contain a combination of the three remaining choices or it could only include sand. Cinder block or CCDD will be blended with sand to reduce the available pore space. The total capacity of the underground structures (basements) for placement of fill is approximately 6 million cubic feet.
There are seven buildings (Figure 1) that will have residual structures beginning three feet below grade. Contaminated concrete from inside the liner in the Containment Buildings will be removed and this will substantially decrease the amount of contamination in the Containment Buildings. Characterization data indicates there is no significant liner contamination or concrete activation past the liner, leaving the Auxiliary Building with the highest residual contamination.
Low-levels of contamination were found in the Turbine Building. The below grade concrete to remain in the Fuel Handling Building and Transfer Canals has not yet been characterized.
2.2 Modeling Overview The Disposal Unit Source Term - Multiple Species (DUST-MS) computer code has been selected to calculate the source term release and equilibrium water concentration at the receptor well which is assumed to be in the center of the backfilled building. DUST-MS has received wide-spread use in subsurface radionuclide release calculations and undergone model validation studies (Sullivan, 1993; 2006). The equilibrium model can be easily calculated by hand.
However, DUST-MS is necessary when simulating diffusion controlled release or transport to a receptor well. To maintain consistency between all calculations DUST-MS was used for all simulations.
An important parameter is the volume of water available to mix with released radionuclides.
Another important parameter defines how the release of contaminants will be modeled. In many 3
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TSD 14-009 Revision 1 buildings the contamination is expected to be loosely bound or near the surface of the remaining structure. In these buildings, the release is assumed to occur instantly, such that the entire inventory is available immediately after license termination. In some buildings the contamination is expected to have diffused into the concrete resulting in volumetrically contaminated concrete. For these buildings, a diffusion controlled release model is used. The Auxiliary Building has been characterized and shown to be contaminated to a depth of at least the first inch of the concrete. The concrete in the Fuel Handling Building and Transfer Canals is also expected to be volumetrically contaminated below the liner but the extent of this contamination will not be characterized until the liner is removed. Diffusion controlled release is assumed for the Auxiliary and Fuel Handling Building/Transfer Canals.
Table 1 summarizes the total fill volume available for mixing and the release assumptions for each building. The mixing volume is calculated assuming that the water level in the basements is equal to the natural water table elevation outside of the basements (i.e., 579 feet), which is the minimum long term level that could exist in the basements. The amount of water available for mixing will be the total fill volume multiplied by the porosity of the backfill. For conservatism it was assumed that the backfill had only 25% porosity. This is believed to be a minimum value for porosity because it will be difficult to achieve this packing density. For example, the native sand has total porosity greater than 30%.
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TSD 14-009 Revision 1 Figure 1 Zion Site building layout.
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TSD 14-009 Revision 1 Table 1 Mixing volume and release rate assumption Building Volume* (m3) Release Rate Assumption Instant Release (loose surface Unit 1 Containment 6.54E+03 contamination)
Instant Release (loose surface Unit 2 Containment 6.54E+03 contamination)
Diffusion Controlled Release (concrete Auxiliary 2.84E+04 contamination at depth in concrete)
Instant Release (the limited 2.61E+04 contamination present is at the Turbine concrete surface with very limited contamination at depth.)
Instant Release (limited or no surface Crib House and Forebay 3.05E+04 contamination)
Waste Water Treatment Instant Release (limited or no surface 1.44E+02 Facility contamination)
Diffusion Controlled Release Spent Fuel Pool and Transfer 2.08E+02 (Concrete contamination expected at Canals depth under the liner)
- (From Farr, 2014)
In the Containment Buildings only loose surface contamination is expected to remain. The distribution of the surface source term is generally expected to be uniform over the remaining liner surface. The release mechanism is therefore Instant Release (e.g. 100% of the inventory is assumed to be instantly released) because the source term is surface contamination only on the remaining steel liner.
The contamination in the Auxiliary Basement is found at depth in the concrete, predominantly in the floor. Diffusion Controlled Release was therefore used to estimate the rate of radionuclide release for the Auxiliary Basement.
The Turbine Basement source term is very limited and associated with surface contamination in concrete and embedded piping in the Turbine Building foundation. The inventory in the concrete and embedded piping is assumed to be instantly released.
There is very little, if any, contamination in the Crib House/Forebay and Waste Water Treatment Facility. The minimal contamination present is assumed to be on the concrete surfaces and instantly released.
Diffusion Controlled Release was used to estimate the source term release rate for the Fuel Handling Building Basement and Fuel Transfer Canals due to expected contamination at depth in concrete after the liners are removed.
In addition to the primary modeling used for 10 CFR 20 Subpart E compliance, a check calculation was performed to determine the water concentration in a well assumed to be placed outside of the building basements at the downstream (eastern) edge of the Turbine Building. The 6
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TSD 14-009 Revision 1 check calculation applies transport modeling to confirm the expectation that the concentration in water outside of the Basements would be lower than inside and that assuming the well is placed inside the Basements is conservative for dose assessment. The area for flow was calculated using the width of the building perpendicular to the primary direction of water flow (from west to east to the Lake in Figure 1) and the mixing height. The contaminated zone in the flow model is the fill material. Outside of the contaminated zone (i.e., outside of the basements) a mixture of fill sand and native soil is simulated. Table 2 contains flow areas for the calculations.
The inventory for each building was based on a uniform contamination level of 1 pCi/m 2 on the wall and floor surfaces. This contamination level was used for modeling convenience only. The total inventory used in the simulation is the value of interest because the total inventory will be used for scaling with the final inventory measured in each basement after remediation is completed. For example, the Auxiliary Building has 6503 m2 of total wall and floor surface area that leads to a total of 6503 pCi in this simulation. To scale to the actual inventory obtained by measurement after remediation is completed, the results of the simulations presented in this report should be multiplied by the ratio of the measured inventory to simulated inventory.
Material properties were chosen to match site-specific values to the extent possible. Sorption coefficient, Kd, values were based on the measured values for Zion soils, concrete, cinder block, and grout (Yim, 2012, Milian, 2014) when available and literature values when site-specific values were not available. A review of literature values and rationale for selecting Kd for dose assessment was performed (Sullivan, 2014). The Kd values selected from the literature were chosen to give a conservative estimate of water concentration (highest value) for dose assessment. When site-specific values are available, the lowest Kd value measured in any fill material or soil was selected.
The compliance assessment requires prediction of the release and transport of contaminants to the hypothetical individual. Characterization studies and assessments by ZionSolutions have identified the following ROCs (Table 3). All nuclides in Table 3 were used in the simulation of maximum groundwater concentration.
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TSD 14-009 Revision 1 Figure 2. Geometry of the Auxiliary Building.
Table 2 Geometric Parameters and Unit Inventory for Residual Structures (Farr, 2014)
Structure Basement Distance Total Floor to Water Inventory Structure Surface Elevation Table Area (Ci)
(feet) meters (m2)
Auxiliary Building 542 11.28 6503 6.50E-09 Unit 1 Containment 565 4.27 2759 2.76E-09 Unit 2 Containment 565 4.27 2759 2.76E-09 Crib House & Forebay 537 12.80 6940 6.94E-09 Turbine Building, Main Steam, Diesel 560 5.79 14679 1.468E-08 Gen Oil Storage Spent Fuel Pool and Transfer Canals 576 0.91 780 7.80E-10 Waste Water Treatment Facility 577 0.61 1124 1.124E-09 8
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TSD 14-009 Revision 1 Table 3 Potential Radionuclides of Concern at the Zion Nuclear Power Station Radionuclides H-3 Co-60 Ni-63 Sr-90 Cs-134 Cs-137 Eu-152 Eu-154 2.3 Release Models 2.3.1 Instant Release For the instant release model the key parameters are the distribution coefficient (Kd), porosity and bulk density of the fill material. The Containment Buildings, Crib House/Forebay, Turbine Building, and the Waste Water Treatment Facility (WWTF) are modeled using an instant release.
2.3.2 Release Rate: Diffusion Controlled Release from the concrete In two of the buildings, Auxiliary and Fuel, there is volumetric contamination in the concrete floors and walls that will release over time as the nuclides diffuse out from the concrete into the water. Therefore, the time-dependent diffusion controlled release rates are used to calculate the maximum water concentrations for the Auxiliary and Fuel Buildings.
Studies have been conducted for the diffusion in concrete of the radionuclides under consideration at Zion (H-3, Co-60, Ni-63, Sr-90, Cs-134, Cs-137, Eu-152, and Eu-154). The diffusion coefficient from concrete will depend on the water to cement ratio used in forming the concrete and the aggregate. A typical range from the literature is presented in Table 4. The maximum in the range was selected for use in the analysis.
Table 4 Typical diffusion coefficients in cement for radionuclides of concern Nuclide Diffusion Coefficient Selected Diffusion Reference Range (cm2/s) Coefficient (cm2/s)
H-3 6.0E 5.5E-07 5.5E-07 Szanto, 2002 Co-60 5.0E 4.1E-11 4.1E-11 Muurinen,1982 Ni-63 8.7E 1.1E-09 1.1E-09 Jakob, 1999 Sr-90 1.0E 5.2E-10 5.2E-10 Sullivan, 1988 Cs-134; Cs-137 4.0E 3.0E-09 3.0E-09 Atkinson, 1986 Eu-152; Eu-154 1.0E 5.0E-11 5.0E-11 Serne, 1992; Serne, 2001 9
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TSD 14-009 Revision 1 In the conceptual model for diffusion controlled release it is assumed that the concrete is uniformly contaminated over a 0.5 inch thickness and that all of the material is released at the surface (i.e. it does not diffuse further into the concrete). This assumption is equivalent to having one side of the contaminated zone as a no flow boundary. In practice, some of the nuclides would continue to diffuse deeper into the concrete initially and thereby increase the time before being released to the water. The assumption that everything is released into the water is modeled with an analytical solution for diffusion from a slab. To simulate release at the surface, the slab is modeled as being one inch thick and allowed to flow out of both sides of the slab.
Using the principle of symmetry, the centerline is a no flow boundary and this is equivalent to having a slab 0.5 inch thick but preventing diffusion further into the cement. This is accomplished in DUST-MS by modeling a slab with a thickness of one inch, which reduces the calculated waste form concentrations from the assumed inventory by a factor of 2 as compared to a one inch thickness. The contributions from both sides of the slab are then summed to calculate the maximum release from one surface of the 0.5 inch slab. Using symmetry, the release from this model, which has two sides, is equivalent to release from a 0.5 inch thick contaminated zone.
2.4 Receptor Well Outside the Turbine Building If CCDD or crushed cinder block is used as fill material, the pH of the water in the fill region will rise to levels that make it non potable. Notwithstanding the high pH condition, the conceptual model assumes that this water will be used as a residential water supply, livestock water supply and for irrigation. This section addresses a more credible scenario where the well is located outside of the basements.
The Auxiliary Building will have the highest levels of residual contamination. The Auxiliary Building is adjacent to the Turbine Building and there are penetrations that will remain in place and connect these buildings. The Containment Buildings are also connected to the Auxiliary Building by penetrations but Containment will have minimal contamination after removal of all internal concrete.
The closest place to put a well in the shallow aquifer outside of the Auxiliary Building is just outside and to the east of the Turbine Building. The Auxiliary Building foundation rests on the clay aquitard and a well located directly to the east of the Auxiliary building, and under the Turbine Building floor would not flow. To examine the maximum concentration that could be obtained from a well in the soil, DUST-MS was used to predict the concentrations 2 meters outside of the eastern edge of the Turbine Building, Figure 1. Therefore, the modeled domain contains the Auxiliary Building and the section of the Turbine Building that aligns with the Auxiliary Building and groundwater flow direction. A schematic representation of the model domain is presented in Figure 3. The dotted rectangular region is the modeled region and consists of clean soil upstream from the Auxiliary Building, the Auxiliary and Turbine Buildings and clean soil downstream of the Turbine Building. A hypothetical well located 2 m from the edge of the Turbine Building is shown. To address the higher contamination levels anticipated in the Auxiliary Building, the Turbine Building contamination level was reduced by a factor of 1000 to 0.001 pCi/m2. The groundwater flow rate through the buildings is assumed to be at the rate determined by the local flow conditions at the site.
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TSD 14-009 Revision 1 Figure 3 Schematic Representation of Flow the geometry used to assess flow to a well outside the Turbine Building.
In RAI 21, the Nuclear Regulatory Commission asked for additional information on the possibility of contributions coming from piping in the Containment Building that connects to the Auxiliary Building. These calculations are provided in Appendix B.
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TSD 14-009 Revision 1
- 3. Analysis Parameters All release models are established using the unit source term and grounded in conservative estimates of site-specific measured values for the model parameters where available. The instant release model was used in buildings with minimal inventory or with only surface contamination expected. The instant release model is meant to provide a conservative upper bound estimate for groundwater concentration. A diffusion release model is used in buildings with volumetric contamination of the concrete.
3.1 Parameters Initial conditions assumed that the groundwater concentration of each contaminant was zero everywhere. The source term is modeled such that the results can be scaled to the actual inventory of the various buildings on site. For this modeling scenario, each building was modeled with the assumption of uniform contamination across the floor of the entire building.
The exact constitution of the backfill has not been decided yet. Therefore, the bulk density and porosity are unknown. A bulk density of 1.5 grams per cubic centimeter (g/cm3) and an effective porosity of 0.25 were selected for the screening model. With any of the fill materials it is difficult to conceive of reducing the packing material below this value. The effective porosity helps determine the amount of water available for mixing and through selecting a low value for this parameter the estimates of concentration in the water will be biased high (e.g. conservative with respect to dose estimates).
The distribution coefficients (Kd) are important parameters in controlling the equilibrium concentrations and transport (if modeled). A study (Sullivan, 2014) reviewed the literature and site-specific data to provide conservative values for Kd in assessing groundwater dose. In selecting values from the literature, environmental conditions with high pH (cement sorption data) as well as environmental data (soil sorption) data were considered. For conservatism the minimum value from these conditions was selected. For nuclides with measured site-specific Kd values, the lowest measured Kd in any backfill or soil was selected. Selected values are in Table 5.
For the base case model it is assumed that there is no flow through the system. This leads to the highest concentrations possible and is conservative. To accomplish this in DUST-MS the flow velocity is set to zero.
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TSD 14-009 Revision 1 Table 5 Selected distribution coefficients (Sullivan, 2014)
Basement Half Fill Kd to Life Be Used Radionuclide (years) cm3/g H-3 12.3 0 Co-60 5.27 223 Ni-63 96 62 Sr-90 29.1 2.3 Cs-134 2.06 45 Cs-137 30 45 Eu-152 13.4 95 Eu-154 8.2 95 3.1.1 Diffusion Controlled Release Model For the diffusion release model the selected diffusion coefficients were presented in Table 4.
The base case model assumes that contamination is uniformly distributed over 0.5 inch in the concrete and all contamination migrates out of the concrete into solution. Additional diffusion into the concrete is not allowed in the model. This maximizes the release rate.
3.1.2 Model Geometry DUST-MS is a one dimensional model. The conceptual model contains a contaminated floor in the direction of flow. DUST-MS model requires a flow area to calculate the correct concentrations above the floor. The flow area is defined as the area perpendicular to the transport direction. In these simulations, the transport direction is towards the Lake. Therefore, the flow is the product of the height of the water table above the floor and the width of the building that is parallel to the Lake. Table 6 provides the height to the water table based on a 579 foot elevation, effective distance parallel to the Lake, flow area, and effective length of the contaminated zone. The product of the flow area and length of the contaminated zone gives the total volume for each building. These widths, height to the water table, and volumes were calculated by ZionSolutions staff (Farr, 2014).
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TSD 14-009 Revision 1 Table 6 Model Geometry for all simulations.
Height Width to Flow or Contaminated Void Space Structure Water Area Radius Zone Length (m) to WT m3 Table (m2) m m
Containment Buildings 20.95 4.27 140.4 44.81 6537 Auxiliary Building 80.11 11.28 903 31.5 28445 Turbine Building 40.84 5.79 571.5 45.73 26135 Crib House and Forebay 52.12 12.8 667.2 45.75 30524 Waste Water Treatment 14.63 0.61 8.919 16.09 144 Facility Spent Fuel Pool and Transfer 10.06 0.91 18.64 11.17 208 Canals 3.1.3 Receptor Well Parameters for Transport Model For the base case the flow velocity is set to zero in the DUST-MS input file. To simulate transport to a receptor well soil properties and the groundwater flow rate are required. These values are presented in Table 7. The Kd values used were identical to those in the equilibrium model. Site-specific soil Kd values for Co (1161 centimeters cubed per gram - cm3/g) and Cs (527 cm3/g) are much higher than used in the analysis and their use would lead to lower predicted concentrations. For conservatism, it was decided that the lowest Kd value from all sources (Sullivan, 2014) would be used. The reason for using the lowest Kd values is that the water leaving the building structures would have a high pH due to the backfill material. This could lead to changes in sorption on the soil materials as compared to the test results obtained using the local groundwater.
Table 7 Transport Parameters used to calculate peak concentrations in a receptor well located outside of the basements.
Parameter Value Reference Soil Density 1.81 (g/cm3) CRA, 2014 Soil Effective Porosity 0.29 CRA, 2014 Groundwater Darcy Velocity 41.6 m/y CRA, 2014 Soil Kd: Co-60 223 (cm3/g) Sullivan, 2014 Ni-63 62 (cm3/g)
Sr-90 2.3 (cm3/g)
Cs-134 45 (cm3/g)
Cs-137 45 (cm3/g)
Eu-152 96 (cm3/g)
Eu-154 95 (cm3/g)
The modeled geometry is presented in Figure 3. The width of the Auxiliary Building is 80.1 m, which is less than the Turbine Building. The one-dimensional simulation requires that the width 14 Page 21 of 56
TSD 14-009 Revision 1 perpendicular to flow remain constant. Therefore, for this simulation only the portions of the Turbine Building downstream from the Auxiliary Building are modeled. The length of the Turbine Building parallel to flow is 29.3 m. Therefore, the total floor area of the Turbine Building for this simulation is 2,344 square meters (m2). This is not the actual area of the Turbine Building modeled in the base case. The receptor well is 2 meters downstream of the Turbine Building. This assumption will have a minor impact on the final results.
The one-dimensional simulation also requires the depth to the water table to remain the same in both buildings. The actual depth to the water table is deeper in the Auxiliary Building as compared to the Turbine Building. The geometry and flow direction requires that any release from the Auxiliary Building travel through the Turbine Building. Therefore, the appropriate depth to the water table for this simulation is that of the Turbine Building, 5.79 m (19 ft.). This value was used to calculate the mixing volume. The total area available for flow (building width multiplied by the height to the water table) is 463.7 m2.
The inventory of the Auxiliary Building is based on 1 pCi/m2 and the total inventory is 2554 pCi.
The inventory of the Turbine Building at the time of license termination will be very close to zero but is assumed to be 0.001 pCi/m2 for a total inventory of 14.7 pCi. The differences in total area lead to the slightly less than a factor of 1,000 difference in total inventory in the two buildings.
3.1.4 Sensitivity Analysis Parameters To quantify the impact of changes in key variables on the predicted concentrations additional calculations were performed. Characterization data indicate that the Auxiliary Building will have the majority of residual contamination. For this reason, all sensitivity analyses will be performed for that building. For sensitivity analysis all parameters were varied by 25% from their initial base case value. The range of parameters is presented in Table 8 Table 8 Parameters and their range in the sensitivity analysis.
Parameter Base Case Value Range Kd Table 6 (nuclide dependent) +/- 25 % of Value in Table 5 Porosity 0.25 0.19 - 0.31 Bulk Density 1.5 g/cm3 1.1 - 1.8 g/cm3 In calculating potential exposures one scenario considers removing the drill spoils from a hypothetical intruder well placed in the middle of the building. These drill spoils are mixed with surface soil and the resulting dose from the contaminated soil is calculated. The Kd values selected for the base case in the backfill were selected to maximize groundwater concentrations.
To examine the impact from using a higher Kd value on the soil concentrations the base case was modified to use the Kd values from the native sand. For tritium (H-3) the Kd value was raised from 0 to 1. Site-specific values for Europium Kd are not available. The 75th percentile value for Kd in soils (7222 ml/g) was used in the analysis (NRC, 2000). Table 9 lists the selected Kd values for the drill spoils sensitivity analysis.
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TSD 14-009 Revision 1 Table 9 Kd values selected to examine the sensitivity of drill spoils predicted soil and groundwater concentrations Nuclide Kd (ml/g)
H-3 1 Co-60 1161 Ni-63 62 Sr-90 2.4 Cs-134 615 Cs-137 615 Eu-152 7721 Eu-154 7721 16 Page 23 of 56
TSD 14-009 Revision 1 4 Results 4.1 Base Case Release Peak Groundwater Concentration Results The conceptual model assumes that the any inventory released instantly comes to equilibrium with the fill material through the sorption process as controlled by the value of Kd. For the instant release model the maximum concentrations occur at time = 0 before any radioactive decay or transport in this model. For the diffusion controlled release, the time to the peak concentration depends on the diffusion coefficient and radionuclide half-life. Tables 9 - 14 provide the maximum concentration in each building. The tables also provide the amount of radioactivity (pCi) in solution, the amount sorbed to the solid material (pCi) and the concentration on the fill material (pCi/g) with a density of 1.5 g/cm3.
4.1.1 Auxiliary Building The base case for the Auxiliary Building assumes a diffusion controlled release. Uniform contamination was assumed over the first 0.5 inch of the concrete. The results of this simulation are provided in Table 10.
Table 10 Auxiliary Building Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2 and diffusion controlled release from 0.5 inch of contaminated concrete. The total inventory for each radionuclide is 6503 pCi.
Peak Peak Diffusion Time to Peak Radioactivity Radioactivity Peak Sorbed Coefficient Kd Peak Concentration in Solution Sorbed Concentration Nuclide (cm2/s) (ml/g) (years) pCi/L pCi pCi pCi/g H-3 5.00E-07 0 0.1 9.10E-04 6503 0.0 0.00E+00 Co-60 4.10E-11 223 4 2.60E-08 0.2 249 5.80E-09 Ni-63 1.10E-09 62 37 1.90E-06 13.6 5051 1.18E-07 Sr-90 5.20E-10 2.3 21 1.96E-05 140.1 1933 4.51E-08 Cs-134 3.00E-09 45 1.5 6.89E-07 4.9 1329 3.10E-08 Cs-137 3.00E-09 45 14 2.47E-06 17.7 4766 1.11E-07 Eu-152 5.00E-11 95 10 1.07E-07 0.8 440 1.03E-08 Eu-154 5.00E-11 95 6 8.38E-08 0.6 341 7.96E-09 Examining Table 10 the impact of diffusion controlled release and sorption is clear. H-3 with no sorption and a high diffusion rate releases almost all the inventory within the first year to solution. Sr-90 with the low Kd value of 2.3 shows slightly more than 4% (140.1 pCi) of the total inventory (6503 pCi) is in solution. For all other nuclides the maximum activity in the water is less than 0.2% of the entire inventory. For Ni-63 the peak activity sorbed to the solid (5051 pCi) is slightly less than 80% of the total activity (6503 pCi). This reflects the time-17 Page 24 of 56
TSD 14-009 Revision 1 dependent release from the concrete and the effects of radioactive decay. The time to peak represents the balance between the release rate, sorption, and radioactive decay. The value in the table is approximate as the concentration shows a broad peak over time. The radionuclides having a short half-life peak the earliest.
4.1.2 Containment Buildings The two Containment Buildings are identical in geometry and therefore, the results for the unit inventory simulation apply to both buildings. In determining the potential dose, the results of this analysis will be scaled by the measured inventory in each building. The Containment Buildings will have all of the concrete inside the liner removed and residual contamination on the liner is assumed to be on the surface. For this reason, the instant release model was used and the results are presented in Table 11.
Table 11 Containment Building Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2. The total inventory for each radionuclide is 2759 pCi.
Half- Peak Radioactivity Radioactivity Sorbed life Kd Concentration in Solution Sorbed Concentration Nuclide (years) (ml/g) pCi/L pCi pCi pCi/g H-3 12.3 0 1.69E-03 2759 0 0 Co-60 5.27 223 1.26E-06 2.1 2756.9 2.81E-07 Ni-63 96 62 4.53E-06 7.4 2751.6 2.81E-07 Sr-90 29.1 2.3 1.14E-04 186.4 2572.6 2.62E-07 Cs-134 2.06 45 6.23E-06 10.2 2748.8 2.80E-07 Cs-137 30 45 6.23E-06 10.2 2748.8 2.80E-07 Eu-152 13.4 95 2.95E-06 4.8 2754.2 2.81E-07 Eu-154 8.2 95 2.95E-06 4.8 2754.2 2.81E-07 For the instant release model more than 99.5% of the material is sorbed on the backfill material for all modeled nuclides except H-3 and Sr-90. Sr-90 with the smallest non-zero Kd value of the group being modeled has slightly less than 7% of the activity in solution. Tritium (H-3), with a value of zero for Kd, has all the activity in solution.
4.1.3 Crib House/Forebay The Crib House/Forebay is expected to contain little or no contamination based on characterization data and the contamination that may be present will be at the surface. For this reason, the instant release model was used. Table 12 provides the results of the analysis.
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TSD 14-009 Revision 1 Table 12 Crib House Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2. The total inventory for each radionuclide is 6940 pCi.
Peak Radioactivity Radioactivity Sorbed Half-life Kd Concentration in Solution Sorbed Concentration Nuclide (years) (ml/g) pCi/L pCi pCi pCi/g H-3 12.3 0 9.08E-04 6936 0.0 1.99E-23 Co-60 5.27 223 6.78E-07 5.2 6930.8 1.51E-07 Ni-63 96 62 2.44E-06 18.6 6917.4 1.51E-07 Sr-90 29.1 2.3 6.14E-05 468.6 6467.4 1.41E-07 Cs-134 2.06 45 3.35E-06 25.6 6910.4 1.51E-07 Cs-137 30 45 3.35E-06 25.6 6910.4 1.51E-07 Eu-152 13.4 95 1.59E-06 12.1 6923.9 1.51E-07 Eu-154 8.2 95 1.59E-06 12.1 6923.9 1.51E-07 4.1.4 Fuel Building The Spent Fuel Pool and Transfer Canals has not been fully characterized at this time. It is believed that there will be volumetric contamination in the concrete below the pool liners. For this reason diffusion controlled release is modeled assuming uniform contamination in the top 0.5 inch of concrete. The results are provided in Table 13.
Table 13 Fuel Building Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2. Release is diffusion controlled from 0.5 inch thick contaminated region. The total inventory for each radionuclide is 780 pCi.
Peak Peak Diffusion Time Peak Radioactivity Radioactivity Peak Sorbed Coefficient Kd to Peak Concentration in Solution Sorbed Concentration Nuclide (cm2/s) (ml/g) (years) pCi/L pCi pCi pCi/g H-3 12.3 0 0.3 1.49E-02 774.8 0 0 Co-60 4.1E-11 223 3.9 4.25E-07 0.02 30 9.48E-08 Ni-63 1.1E-09 72 36 3.13E-05 1.6 605 1.94E-06 Sr-90 5.2E-10 2.3 21 3.21E-04 16.7 230 7.38E-07 Cs-134 3.0E-09 45 1.5 1.13E-05 0.6 159 5.09E-07 Cs-137 3.0E-09 45 13.3 4.07E-05 2.1 571 1.83E-06 Eu-152 5.0E-11 96 9.5 1.75E-06 0.09 52 1.68E-07 Eu-154 5.0E-11 95 6.2 1.37E-06 0.07 41 1.30E-07 The impact of diffusion controlled release on peak concentrations is slightly more pronounced than in the Auxiliary Building with a peak solution concentration for Sr-90 slightly in excess of 2 percent of the total inventory. The H-3 concentration predicted for the Fuel Building (0.015 19 Page 26 of 56
TSD 14-009 Revision 1 pCi/L) is the highest predicted concentration for any of the buildings. This is due to the small amount of water available for mixing and the high diffusion release rate (over 99% of the inventory is released in the first year). The mixing height is only 0.91 m as compared to 11.28 m for the Auxiliary Building.
4.1.5 Turbine Building The Turbine Building is expected to contain little or no contamination based on characterization data and contamination that was identified was predominantly at the surface. For this reason the instant release model is used. The results are provided in Table 14.
Table 14 Turbine Building Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2. The total inventory for each radionuclide is 14679 pCi.
Peak Radioactivity Radioactivity Sorbed Half-life Kd Concentration in Solution Sorbed Concentration Nuclide (years) (ml/g) pCi/L pCi pCi pCi/g H-3 12.3 0 2.25E-03 14679 0.0 0 Co-60 5.27 223 1.68E-06 11.0 14668.0 3.74E-07 Ni-63 96 62 6.02E-06 39.4 14639.6 3.73E-07 Sr-90 29.1 2.3 1.52E-04 991.8 13687.2 3.49E-07 Cs-134 2.06 45 8.29E-06 54.2 14624.8 3.73E-07 Cs-137 30 45 8.29E-06 54.2 14624.8 3.73E-07 Eu-152 13.4 95 3.93E-06 25.4 14653.6 3.74E-07 Eu-154 8.2 95 3.93E-06 25.7 14653.3 3.74E-07 Similar to the Crib House building, Sr-90 shows the highest solution concentration for sorbing nuclides and 6.7% of the Sr-90 is in the groundwater. Tritium (H-3) which does not sorb has the highest solution concentration.
4.1.6 Waste Water Treatment Facility The WWTF is expected to contain little or no contamination based on characterization data and any contamination that may be present would be on the surface. For this reason the instant release model is used. The results are provided in Table 15.
The Waste Water Treatment Facility shows the highest peak concentrations per unit source term of all of the buildings with the exception of H-3. The cause for this is the very low mixing volume which is 143 m3 and high surface area 1124 m2. The surface area to volume ratio for this building is 7.8 m-1, the largest of any building with an instant release source term. The inventory is directly proportional to surface area. Therefore, a high surface area to volume ratio will produce higher peak concentrations. The Fuel Building has a higher surface area to volume ratio but release was controlled by diffusion which limited the concentrations of everything except H-3 to lower levels than in the Waste Water Treatment Facility.
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TSD 14-009 Revision 1 Table 15 Waste Water Treatment Facility Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2. The total inventory for each radionuclide is 1124 pCi.
Peak Radioactivity Radioactivity Sorbed Half-life Kd Concentration in Solution Sorbed Concentration Nuclide (years) (ml/g) pCi/L pCi pCi pCi/g H-3 12.3 0 3.13E-02 1124 0.0 0 Co-60 5.27 223 2.34E-05 0.8 1123.2 5.22E-06 Ni-63 96 62 8.40E-05 3.0 1121.0 5.21E-06 Sr-90 29.1 2.3 2.12E-03 75.9 1048.1 4.87E-06 Cs-134 2.06 45 1.16E-04 4.1 1119.9 5.20E-06 Cs-137 30 45 1.16E-04 4.1 1119.9 5.20E-06 Eu-152 13.4 95 5.43E-05 1.9 1122.1 5.21E-06 Eu-154 8.2 95 5.48E-05 2.0 1122.0 5.21E-06 4.2 Sensitivity Analysis A sensitivity analysis was performed on the key parameters in the base case model for the Auxiliary Building. The key parameters in the base case model are the distribution coefficient Kd, porosity, and bulk density. Each of these was varied as defined in Table 8 for a total of six test cases. Appendix A contains the detailed results of these simulations and includes Tables identical in form to Tables 10 - 15 with the peak concentration, amount of activity in solution and sorbed to the solid, and the activity concentration on the solid (pCi/g). Additionally, there is a table providing the percent (%) change due to the variation in the parameter from the base case. The % Change was defined as:
% Change = 100*(Sensitivity Case - Base Case)/Base Case.
Thus, the % Change is positive if the sensitivity case value exceeds the base case value.
The major findings of the sensitivity analyses are:
For all nuclides except H-3, most of the activity is sorbed onto the backfill material.
Strontium with the lowest Kd still had more than 90% of the activity sorbed on the backfill.
Kd: An increase in Kd caused a decrease in solution concentration and a slight increase in sorbed concentration. Solution concentration is approximately inversely proportional to Kd. The 25% change in Kd had a minimal impact on the amount sorbed or the backfill concentration (pCi/g). Strontium showed the largest percentage change in sorbed concentration of all the nuclides but it was less than 2.5%.
Porosity: Changing porosity had a minor impact on the amount sorbed and solution concentration. The amount of radioactivity in solution was proportional to the porosity.
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TSD 14-009 Revision 1 This reflects the availability of water with higher porosity having more water available for mixing and a higher total amount of activity in the water.
Density: The solution concentration, sorbed concentration and amount in solution are inversely proportional to density. Increasing density causes a decrease in solution concentration. The change in density has a minor impact (< 2%) on the total amount of radioactivity that is sorbed.
4.2.1 Sensitivity to Release Rate The base case model for the Auxiliary Building assumes diffusion controlled release from a 0.5 inch thick contaminated zone. For sensitivity analysis release was simulated from a 1 inch and 2 inch thick contaminated zone. In all cases, the total inventory for each nuclide remained constant at 6503 pCi. Changes in the depth of contamination can lead to changes in the total amount of mass released, the peak concentration, and the time to reach the peak concentration.
Table 16 examines the impact of contaminated zone thickness on the percentage of the total inventory released into solution over time and compares the change in total mass released to the base case 1/2 inch thick contaminated zone. H-3 has the highest diffusion coefficient and releases over 98% of the inventory in all three cases and therefore, the contaminated zone thickness only has a minor impact on the total mass released. The nuclides with a short half-life or a low diffusion coefficient in this simulation (Co-60, Sr-90, Cs-134, Eu-152, and Eu-154) show similar behavior and increasing the contaminated zone thickness by a factor of two leads to a factor of two decrease in the amount of mass released. Thus, in this region, the mass release is almost directly proportional to the contaminated zone thickness for these nuclides. The longer lived nuclides with the higher diffusion coefficients (Cs-137, and Ni-63) show similar trends but the response is much further from linear with distance than the shorter lived nuclides.
Table 16 Comparison of the percentage of the total inventory released based on the thickness of the contaminated zone. Thicknesses analyzed were 0.5 inch (base case), 1 inch and 2 inch.
0.5 inch 1 inch 2 inch thick thick thick Diffusion % Mass % Mass % % Mass %
Nuclide Coefficient Released Released change Released change H-3 5.5E-07 100.0 99.7 -0.3 98.2 -1.8 Co-60 4.1E-11 7.9 4.0 -49.8 2.0 -74.4 Ni-63 1.1E-09 92.2 74.8 -18.9 43.3 -53.0 Sr-90 5.2E-10 61.9 32.9 -46.8 16.7 -72.9 Cs-134 3.0E-09 42.4 21.4 -49.6 10.9 -74.4 Cs-137 3.0E-09 90.8 71.0 -21.8 40.9 -54.9 Eu-152 5.0E-11 13.8 6.9 -49.7 3.5 -74.4 Eu-154 5.0E-11 10.8 5.4 -49.7 2.8 -74.4 22 Page 29 of 56
TSD 14-009 Revision 1 Table 17 provides the peak water concentration as a function of contaminated zone thickness and the percentage change from the base case (0.5 inch thick contaminated zone). The peak concentrations followed the same trends as the percentage of total mass released. H-3 showed only a minor decrease as most of the mass is released quickly for contaminated thickness of less than 2 inches. The other nuclides showed an almost linear response with contamination thickness as increasing the thickness by a factor of 2 leading to a decrease in peak concentration by a factor of 2.
Table 17 Comparison of the peak water concentration based on the thickness of the contaminated zone. Thicknesses analyzed were 1 inch (base case), 1/2 and 2 inch.
0.5 inch thick 1 inch thick 2 inch thick Peak Peak Peak Diffusion concentration concentration % concentration %
Nuclide Coefficient (pCi/L) (pCi/L) change (pCi/L) change H-3 5.5E-07 9.10E-04 9.00E-04 -1.1 8.57E-04 -5.8 Co-60 4.1E-11 2.60E-08 1.30E-08 -50.0 6.64E-09 -74.5 Ni-63 1.1E-09 1.90E-06 1.05E-06 -44.7 5.37E-07 -71.7 Sr-90 5.2E-10 1.96E-05 9.84E-06 -49.8 5.01E-06 -74.4 Cs-134 3.0E-09 6.89E-07 3.41E-07 -50.5 1.76E-07 -74.5 Cs-137 3.0E-09 2.47E-06 1.32E-06 -46.6 6.7E-07 -72.9 Eu-152 5.0E-11 1.07E-07 5.38E-08 -49.7 2.74E-08 -74.4 Eu-154 5.0E-11 8.38E-08 4.21E-08 -49.8 2.14E-08 -74.5 Table 18 provides the time to reach the peak concentration as a function of contaminated zone thickness and the percentage change from the base case (0.5 inch thick contaminated zone). The time to reach the peak concentration is a balance between the diffusion release rate and the radioactive decay rate. H-3 is very sensitive to contaminated zone thickness in the time to reach the peak concentration. This is because of the high release rate (high diffusion coefficient) of H-3. The short-lived species (Co-60, Sr-90, Cs-134, Eu-152, and Eu-154) show no sensitivity to the peak concentration time for any of the contaminated zone thicknesses tests. Ni-63 showed moderate sensitivity with the time to reach peak concentration varying between 37 and 72 years.
Cs-137 showed an increase in the time to reach peak concentration of 57% in going to the 1 inch thick contaminated zone from the base case. However, it did not show a change in the time to reach the peak concentration above 1 inch contaminated zone thickness.
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TSD 14-009 Revision 1 Table 18 Comparison of the time to reach the peak concentration in solution based on the thickness of the contaminated zone. Thicknesses analyzed were 0.5 inch (base case), 1 inch and 2 inch.
0.5 inch thick 1inch thick 2 inch thick Peak Peak Peak Diffusion concentration concentration % concentration %
Nuclide Coefficient time (yrs) time (yrs) change time (yrs) change H-3 5.5E-07 0.1 0.3 200.0 1.1 1000.0 Co-60 4.1E-11 4 4 0.0 4 0.0 Ni-63 1.1E-09 37 63 70.3 72 94.6 Sr-90 5.2E-10 21 21 0.0 21 0.0 Cs-134 3.0E-09 1.5 1.5 0.0 1.5 0.0 Cs-137 3.0E-09 14 22 57.1 22 57.1 Eu-152 5.0E-11 10 10 0.0 10 0.0 Eu-154 5.0E-11 6 6 0.0 6 0.0 4.2.2 Drill Spoils Sensitivity to Kd As discussed in section 3.1.3 one exposure scenario includes using the drill spoils and mixes them with the native soil. To examine the change in drill spoils radionuclide concentration the Kd values in Table 9 were used. Table 19 provides the results for the new Kd values in the Auxiliary Building with all other parameters unchanged.
Table 19 Sensitivity of Drill Spoils to Distribution Coefficient (Kd)
Base Drill Case Spoils Peak Radioactivity Radioactivity Sorbed Kd Kd Concentration in Solution Sorbed Concentration Nuclide (ml/g) (ml/g) pCi/L pCi pCi pCi/g H-3 0 1 1.28E-04 914.7 5488 1.28E-07 Co-60 223 1161 4.99E-09 0.04 248 5.80E-09 Ni-63 62 62 1.90E-06 13.6 5051 1.18E-07 Sr-90 2.3 2.3 1.96E-05 140.1 1933 4.51E-08 Cs-134 45 615 5.05E-08 0.4 1332 3.11E-08 Cs-137 45 615 1.82E-07 1.3 4799 1.12E-07 Eu-152 96 7221 1.41E-09 0.0 437 1.03E-08 Eu-154 96 7221 1.10E-09 0.0 341 7.96E-09 Table 20 compares the sensitivity case to the base case for the peak concentration and peak sorbed concentration. The results for Ni-63 and Sr-90 are identical as the Kd values are the same in the two simulations. For the other nuclides increasing the Kd value led to lower predicted 24 Page 31 of 56
TSD 14-009 Revision 1 Table 20 Comparison of Base Case and Drill Spoils case Base Drill Base Case: Drill Spoils: Base Case: Drill Spoils:
Case Spoils Peak Peak Sorbed Sorbed Kd Kd Concentration Concentration Concentration Concentration Nuclide (ml/g) (ml/g) pCi/L pCi/L pCi/g pCi/g H-3 0 1 9.10E-04 1.30E-04 0.00E+00 1.30E-07 Co-60 223 1161 2.60E-08 4.99E-09 5.80E-09 5.80E-09 Ni-63 62 62 1.90E-06 1.90E-06 1.18E-07 1.18E-07 Sr-90 2.3 2.3 1.96E-05 1.96E-05 4.51E-08 4.51E-08 Cs-134 45 615 6.89E-07 5.05E-08 3.10E-08 3.11E-08 Cs-137 45 615 2.47E-06 1.82E-07 1.11E-07 1.11E-07 Eu-152 96 7221 1.07E-07 1.41E-09 1.03E-08 1.03E-08 Eu-154 96 7221 8.38E-08 1.11E-09 7.96E-09 8.02E-09 peak groundwater concentrations. This is most apparent for H-3 where the base case Kd value is 0 ml/g. The interesting point about this table is that even with a factor of ten increase in Kd (for example, Cs and Eu) the sorbed concentration increased only slightly (< 2%). This is a reflection of the fact that for Kd values greater than 10 more than 99% of the mass released is sorbed and therefore increasing Kd further has only a minor impact on the sorbed concentration.
4.3 Outside Receptor Well Concentration in Transport Model The time evolution of concentration at a receptor well located two meters outside the Turbine Building was simulated using the backfill material Kd values in Table 4, the soil Kd and groundwater parameters in Table 7, and the geometry in Figure 3. The initial contamination level in the Auxiliary Building (1 pCi/m2) was conservatively assumed to be 1000 times greater than in the Turbine Building (0.001 pCi/m2). This assumption led to a total inventory of 6503 pCi in the Auxiliary Building and 14.7 pCi in the Turbine Building. Consistent with the Base Case, diffusion controlled release is assumed for the Auxiliary Building and Instant Release is assumed for the Turbine Building.
Table 18 provides the peak concentration in the Auxiliary Building, Turbine Building, Edge of the Turbine Building, and the Receptor Well. To quantitatively define the reduction in concentration from the Auxiliary Building to the Receptor Well the ratio of peak concentration at the well to the peak concentration in the Auxiliary Building is provided. The time to reach the peak at the Receptor Well is also provided. Recalling that the initial inventory in the Turbine Building was 450 times lower than in the Auxiliary Building, it is clear that Co-60 and Cs-134 did not move from the Auxiliary Building to the receptor well in any appreciable quantities. For the shorter lived nuclides (Co-60, Cs-134, Eu-152, and Eu-154) the combination of radioactive decay and sorption reduced the concentration by around a factor of ten in traveling two meters from the edge of the Turbine Building to the Receptor Well. H-3, the most mobile nuclide reached a maximum at the well after 1.5 years and showed a peak concentration ratio of 0.8 thus the transport through the Turbine Building did little to diminish the concentration of H-3. Sr-90, which exhibits some sorption but has a longer half-life than H-3, had a peak concentration ratio 25 Page 32 of 56
TSD 14-009 Revision 1 of 0.78 after 23 years, slightly less than that for the more mobile H-3. All other nuclides had a peak concentration ratio of less than 2%.
Table 21 Comparison of Peak Concentrations in the modeled region.
Edge of Ratio Turbine Turbine Receptor Well to Time to Aux Bldg. Bldg. Bldg. Well Auxiliary peak (pCi/L) (pCi/L) (pCi/L) (pCi/L) Building (years)
H-3 1.48E-03 1.48E-03 1.21E-03 1.19E-03 0.80 1.5 Co-60 2.5E-08 2.1E-09 2.1E-09 2.7E-11 0.001 15 Ni-63 2.02E-06 6.38E-07 5.23E-08 3.5E-08 0.017 >300 Sr-90 1.10E-05 1.18E-05 8.81E-06 8.60E-06 0.78 23 Cs-134 6.74E-07 1.02E-08 1.00E-08 3.93E-10 0.001 4.5 Cs-137 2.56E-06 1.80E-07 1.01E-08 4.61E-09 0.002 21 Eu-152 1.04E-07 4.8E-09 4.76E-09 6.94E-10 0.007 18 Eu-154 8.20E-08 4.85E-09 4.81E-09 4.30E-10 0.005 13 4.4 Discussion The simulation of a well located in the middle of the contaminated zone is intended to provide a reasonable upper bound on peak contaminant concentrations. The following qualitative arguments support this assertion.
The Reasonably Foreseeable Scenario, defined in NUREG 1757 as a land use scenario that is likely within the next 100 years, would not include an onsite water well which is prohibited by local municipal code.
If the local laws were ignored, it is unlikely that anyone would drill through the backfill (concrete construction debris) to install a well.
If a well was installed, the water will be non-potable due to the high pH (>10) that will occur from leaching of the concrete construction debris.
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TSD 14-009 Revision 1 5 Validation The instant release model reduces to a simple mixing bath model where the entire inventory is at equilibrium with the backfill material. The concentration for this model can be calculated as:
C = M/(V*(1+
- Kd// )
Where C= concentration in solution (pCi/L)
M = inventory (pL)
V = volume (L) (2.65E7 L in Turbine Building).
= effective porosity (0.25)
= bulk density (g/cm3) (1.5 g/cm3)
Kd = distribution coefficient (cm3/g)
A comparison was made between the DUST-MS output and the analytical solution in the equation above for the Turbine Building as an example of an instant release basement. The results showed an excellent match between the two predictions, Table 19.
Table 22 Comparison between Analytical Solution and DUST-MS results for the Turbine Building.
DUST-MS Nuclide Kd C (pCi/L) C(pCi/L)
H-3 0 2.21E-03 2.21E-03 Co-60 223 1.65E-06 1.65E-06 Ni-63 62 5.94E-06 5.94E-06 Sr-90 2.3 1.50E-04 1.50E-04 Cs-134 45 8.17E-06 8.17E-06 Cs-137 45 8.17E-06 8.17E-06 Eu-152 95 3.88E-06 3.88E-06 Eu-154 95 3.88E-06 3.88E-06 Similar calculations were performed for all buildings and showed a good match between the two models.
27 Page 34 of 56
TSD 14-009 Revision 1 6 Conclusions A model for predicting peak groundwater concentrations at the ZSRP Site after decommissioning has been developed. The model uses the DUST-MS simulation model which calculates the release and transport of radioactive contamination in a groundwater system. The analysis is based on a unit source term of 1 pCi/m2 on the entire wall and floor surface area of each of the seven buildings that will have a residual below ground, backfilled structure. Conservative assumptions based on existing data were used in the screening model for selecting parameters that impact groundwater concentration (Kd, porosity, bulk density, no flow). For example, the Kd value selected for the fill material was the lowest measured value using site-specific groundwater for any soil or fill material. The results of the model can be combined with measured data after characterization is completed to determine peak groundwater dose for all the nuclides.
A sensitivity analysis was performed for the key variables (Kd, effective porosity, bulk density) for the Auxiliary Building base case. The results of the analysis showed that the peak water concentration was inversely proportional to bulk density and Kd. The solution concentration was weakly sensitive to changes in porosity. In all cases, more than 90% of the nuclide inventory is sorbed onto the fill material.
A sensitivity analysis was performed on the release model through comparison of the diffusion change in total mass released, peak concentration, and time to reach the peak concentration for the base case, one inch contaminated zone, to results from simulations with one-half and two inch contaminated zone. For H-3, which has the highest diffusion coefficient, the mass released and peak concentration were not sensitive to the length of the contaminated zone. Over 98% of the mass was released in all three simulations. The other nuclides showed close to an inverse linear dependence on contaminated zone length with the mass release and peak concentration decreasing by close to a factor of two with an increase in length of a factor of two. The time to reach the peak concentration was independent of the length of the contaminated zones for short-lived nuclides (other than H-3) indicating that a balance between release rate and radioactive decay was achieved. For H-3 the high release rate caused the peak concentration to be reached in 0.1 years for the shortest contaminated length (1/2 inch) and 1.4 years for the two inch contaminated length simulation.
Removing the assumption of a well placed in the middle of the fill material and placing the Receptor Well two meters outside the Turbine Building, which is the closest soil (e.g. non-building) location to the Auxiliary Building where the highest residual contamination will remain, led to a three to four order of magnitude reduction in peak concentration for short-lived nuclides (Co-60; Cs-134, Eu-152, and Eu-154), a two order of magnitude reduction for Cs-137, and a factor of fifty reduction for Ni-63. H-3 showed a 20% reduction in peak dose due radioactive decay and transport to the well. Sr-90, which has high mobility and longer half-life than H-3, showed a 22% reduction in peak concentration at the Receptor Well as compared to in the Auxiliary Building.
28 Page 35 of 56
TSD 14-009 Revision 1 7 References Atkinson, A., Nelson, K., and Valentine, T.M., Leach test characterization of cement-based nuclear waste forms, Nuclear and Chemical Waste Management, Vol. 6 (1986), 241 - 253.
Conestoga-Rovers & Associates, 2014, 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.
Farr, H.C., Re: New Volumes e-mail 9/24/14 to T. Sullivan Jakob, A., F.-A. Sarott and P. Spieler, "Diffusion and sorption on hardened cement pastes - experiments and modeling results", Paul Scherer Institute. PSI-Bericht Nr. 99-05 ISSN 1019-0643, August 1999.
Milian, L., T. Sullivan (2014). Sorption (Kd) measurements on Cinder Block and Grout in Support of Dose Assessments for Zion Nuclear Station Decommissioning, Brookhaven National Laboratory Report to ZionSolutions, April 2014.
Muurinnen, A, J. Rantanen, R. Ovaskainen and O.J. Heinonen, Diffusion Measurements in Concrete and Compacted Bentonite, Proceedings of the Materials Research Meeting, 1982.
Serne, R. J., R.O. Lokken, and L.J. Criscenti. Characterization of Grouted LLW to Support Performance Assessment. Waste Management 12: 271-287, 1992.
Serne, J., Selected Diffusion Coefficients for Radionuclides in Cement, personal communication.
Sullivan, T.M., "DUST - Disposal Unit Source Term: Data Input Guide." NUREG/CR-6041, BNL-NUREG-52375, 1993.
Sullivan, T.M., C.R. Kempf, C.J. Suen, and S.F. Mughabghab, "Low-Level Radioactive Waste Source Term Model Development and Testing," NUREG/CR-5204, BNL-NUREG-52 160, Brookhaven National Laboratory, 1988.
Sullivan, T.M., "DUSTMS_D - Disposal Unit Source Term - Multiple Species - Distributed Failure Data Input Guide. Rev 1., BNL-75554-2006, Brookhaven National Laboratory, Upton, NY, 11973, January, 2006.
Sullivan, T.M., Recommended Values for the Distribution Coefficient (Kd) to be Used in Dose Assessments for Decommissioning the Zion Nuclear Power Plant, Revision 1, BNL-Letter Report, September 24, 2014.
Szanto, Zs, Svingor, M. Molnir, L. Palcsu, I. Futo, Z. Szucs. "Diffusion of 3H, 99Tc, 125I, 36Cl, and 85 Sr in granite, concrete and bentonite," Journal of Radioanalytical and 29 Page 36 of 56
TSD 14-009 Revision 1 Nuclear Chemistry, Vol. 252, No. 1 (2002) 133-138.
U.S. Nuclear Regulatory Commission, (NRC, 2000). Development of Probabilistic RESRAD 6.0 and RESRADBUILD 3.0 Computer Codes, NUREG/CR-6697, U.S. Nuclear Regulatory Commission, December 2000.
Yim, S.P, T.M. Sullivan, and L. Milian, Sorption (Kd) measurements in Support of Dose Assessments for Zion Nuclear Station Decommissioning, Brookhaven National Laboratory Report to ZionSolutions, December 12, 2012.
30 Page 37 of 56
TSD 14-009 Revision 1 Appendix A: Sensitivity Analysis Results A.1: Base Case The base case for the Auxiliary Building is diffusion-controlled release from the concrete floors.
The initial inventory for each nuclide was 6503 pCi. There is a major difference between non-sorbing nuclides (H-3) and sorbing nuclides. The non-sorbing nuclide showed approximately 96% of the inventory in solution. With the other 4% decayed prior to release from the floors and wall. The sorbing nuclides had less than 1.2% in solution with most of the released mass sorbed.
Examining the Peak Radioactivity Sorbed shows that less than 1/2 of the total inventory was on the backfill at any time.
Time Peak Peak Diffusion Kd to Peak Radioactivity Peak Sorbed Coefficient Concentration Radioactivity Peak in Solution Sorbed Concentration Nuclide (cm2/s) (ml/g) (years) pCi/L pCi pCi pCi/g H-3 5.00E-07 0 1.5 8.70E-04 6267 0.0 0.00E+00 Co-60 4.10E-11 223 3.8 1.30E-08 0.09 125.3 2.90E-09 Ni-63 1.10E-09 62 72 1.05E-06 7.56 2813.7 6.51E-08 Sr-90 5.20E-10 2.3 22 9.84E-06 70.88 978.2 2.26E-08 Cs-134 3.00E-09 45 1.5 3.41E-07 2.46 663.2 1.53E-08 Cs-137 3.00E-09 45 22 1.32E-06 9.51 2567.3 5.94E-08 Eu-152 5.00E-11 96 9.5 5.38E-08 0.39 223.2 5.16E-09 Eu-154 5.00E-11 95 6 4.21E-08 0.30 172.9 4.00E-09 31 Page 38 of 56
TSD 14-009 Revision 1 A.2: High Kd Kd values are in the table below. They were increased by 25% from the base case value.
Increasing the Kd value increases the amount of sorption and reduces the solution concentration.
For non-sorbing nuclides there is no impact for changes in Kd.
A negative number means that the base case value is greater than the sensitivity case value.
Peak Peak Diffusion Peak Peak Sorbed Kd Radioactivity in Radioactivity Coefficient Concentration Concentration Solution Sorbed Nuclide (cm2/s) (ml/g) pCi/L pCi pCi pCi/g H-3 5.50E-07 0 8.70E-04 6267 Co-60 4.10E-11 278.8 1.04E-08 0.075 125.3 2.90E-09 Ni-63 1.10E-09 77.5 8.38E-07 6.037 2807.0 6.49E-08 Sr-90 5.20E-10 2.88 7.97E-06 57.412 992.1 2.30E-08 Cs-134 3.00E-09 56.3 2.74E-07 1.974 666.7 1.54E-08 Cs-137 3.00E-09 56.3 1.06E-06 7.636 2579.3 5.97E-08 Eu-152 5.00E-11 120 4.30E-08 0.310 223.0 5.16E-09 Eu-154 5.00E-11 118.8 3.37E-08 0.243 173.0 4.00E-09
% Change from the Base case = 100*(Sensitivity Case- Base Case)/Base Case Peak Radioactivity Radioactivity Sorbed Concentration in Solution Sorbed Concentration Nuclide pCi/L pCi pCi pCi/g H-3 0.0 0.0 Co-60 -20.0 -20.0 0.0 0.0 Ni-63 -20.2 -20.2 -0.2 -0.2 Sr-90 -19.0 -19.0 1.4 1.4 Cs-134 -19.6 -19.6 0.5 0.5 Cs-137 -19.7 -19.7 0.5 0.5 Eu-152 -20.1 -20.1 -0.1 -0.1 Eu-154 -20.0 -20.0 0.1 0.1 32 Page 39 of 56
TSD 14-009 Revision 1 A.3: Low Kd Kd values are shown in the table below and were reduced by 25% from the base case values.
Reducing Kd increases the amount in solution for sorbing nuclides but does not impact the total amount sorbed. For non-sorbing nuclides the change in Kd has no impact.
Diffusion Peak Radioactivity Radioactivity Sorbed Kd Coefficient Concentration in Solution Sorbed Concentration Nuclide (cm2/s) (ml/g) pCi/L pCi pCi pCi/g H-3 5.50E-07 0 8.70E-04 6267 Co-60 4.10E-11 167 1.74E-08 0.13 126 2.91E-09 Ni-63 1.10E-09 47 1.39E-06 10.01 2794 6.46E-08 Sr-90 5.20E-10 1.73 1.28E-05 92.20 954 2.21E-08 Cs-134 3.00E-09 34 4.51E-07 3.25 658 1.52E-08 Cs-137 3.00E-09 34 1.74E-06 12.53 2538 5.87E-08 Eu-152 5.00E-11 72 7.17E-08 0.52 223 5.16E-09 Eu-154 5.00E-11 72 5.61E-08 0.40 175 4.04E-09
% Change from the Base case = 100*(Sensitivity Case- Base Case)/Base Case Peak Radioactivity Radioactivity Sorbed Concentration in Solution Sorbed Concentration Nuclide pCi/L pCi pCi pCi/g H-3 0.0 0.0 Co-60 33.8 33.8 0.4 0.4 Ni-63 32.4 32.4 -0.7 -0.7 Sr-90 30.1 30.1 -2.4 -2.4 Cs-134 32.3 32.3 -0.8 -0.8 Cs-137 31.8 31.8 -1.1 -1.1 Eu-152 33.3 33.3 0.0 0.0 Eu-154 33.3 33.3 1.0 1.0 33 Page 40 of 56
TSD 14-009 Revision 1 A.4: High Porosity The porosity was increased to 0.31 from the base case value of 0.25. Increasing porosity did not impact the solution concentration but did increase the amount of radioactivity in solution due to the greater amount of water for sorbing nuclides. For non-sorbing nuclides increasing porosity decreased the solution concentration but did not impact the total amount in solution.
Diffusion Peak Radioactivity Radioactivity Sorbed Kd Coefficient Concentration in Solution Sorbed Concentration Nuclide (cm2/s) (ml/g) pCi/L pCi pCi pCi/g H-3 5.50E-07 0 8.70E-04 6267 Co-60 4.10E-11 223 1.30E-08 0.12 125 2.90E-09 Ni-63 1.10E-09 62 1.05E-06 9.38 2814 6.51E-08 Sr-90 5.20E-10 2.3 9.68E-06 86.47 962 2.23E-08 Cs-134 3.00E-09 45 3.41E-07 3.05 663 1.53E-08 Cs-137 3.00E-09 45 1.32E-06 11.79 2567 5.94E-08 Eu-152 5.00E-11 96 5.38E-08 0.48 223 5.16E-09 Eu-154 5.00E-11 95 4.21E-08 0.38 173 4.00E-09
% Change from the Base case =
100*(Sensitivity Case- Base Case)/Base Case Peak Radioactivity Radioactivity Sorbed Concentration in Solution Sorbed Concentration Nuclide pCi/L pCi pCi pCi/g H-3 -19.3 0.1 Co-60 0.0 24.0 0.0 0.0 Ni-63 0.0 24.0 0.0 0.0 Sr-90 -1.6 22.0 -1.6 -1.6 Cs-134 0.0 24.0 0.0 0.0 Cs-137 0.0 24.0 0.0 0.0 Eu-152 0.0 24.0 0.0 0.0 Eu-154 0.0 24.0 0.0 0.0 34 Page 41 of 56
TSD 14-009 Revision 1 A.5: Low Porosity The porosity was decreased to 0.19 from the base case value of 0.25. For sorbing nuclides decreasing the porosity did not impact the solution concentration but it did reduce the total amount of radioactivity in the water. For non-sorbing nuclides decreasing the porosity increased the solution concentration but did not impact the amount in solution.
Diffusion Peak Radioactivity Radioactivity Sorbed Kd Coefficient Concentration in Solution Sorbed Concentration Nuclide (cm2/s) (ml/g) pCi/L pCi pCi pCi/g H-3 5.50E-07 0 1.14E-03 6241 Co-60 4.10E-11 223 1.30E-08 0.07 125 2.90E-09 Ni-63 1.10E-09 62 1.05E-06 5.7 2814 6.51E-08 Sr-90 5.20E-10 2.3 9.68E-06 53.0 962 2.23E-08 Cs-134 3.00E-09 45 3.41E-07 1.9 663 1.53E-08 Cs-137 3.00E-09 45 1.32E-06 7.2 2567 5.94E-08 Eu-152 5.00E-11 95 5.38E-08 0.3 221 5.11E-09 Eu-154 5.00E-11 96 4.21E-08 0.23 175 4.04E-09
% Change from the Base case =
100*(Sensitivity Case- Base Case)/Base Case Peak Radioactivity Radioactivity Sorbed Concentration in Solution Sorbed Concentration Nuclide pCi/L pCi pCi pCi/g H-3 31.0 -0.4 Co-60 0.0 -24.0 0.0 0.0 Ni-63 0.0 -24.0 0.0 0.0 Sr-90 -1.6 -25.2 -1.6 -1.6 Cs-134 0.0 -24.0 0.0 0.0 Cs-137 0.0 -24.0 0.0 0.0 Eu-152 0.0 -24.0 -1.0 -1.0 Eu-154 0.0 -24.0 1.1 1.1 35 Page 42 of 56
TSD 14-009 Revision 1 A.6: High Bacfkill Density The backfill density was increased to 1.8 g/cm3 from the base case value of 1.5 g/cm3.
Increasing the density caused both the solution concentration and sorbed concentration to decrease for sorbing nuclides. This is because the extra mass provided more sorption to reduce solution concentrations and more mass to sorb onto and therefore lower sorbed concentrations.
The density did not impact non-sorbing nuclides.
Diffusion Peak Radioactivity Radioactivity Sorbed Coefficient Kd Concentration in Solution Sorbed Concentration Nuclide (cm2/s) (ml/g) pCi/L pCi pCi pCi/g H-3 5.50E-07 0 8.70E-04 6267 Co-60 4.10E-11 223 1.09E-08 0.08 126.1 2.43E-09 Ni-63 1.10E-09 62 8.79E-07 6.33 2827 5.45E-08 Sr-90 5.20E-10 2.3 8.29E-06 59.72 989 1.91E-08 Cs-134 3.00E-09 45 2.84E-07 2.05 663 1.28E-08 Cs-137 3.00E-09 45 1.10E-06 7.92 2567 4.95E-08 Eu-152 5.00E-11 96 4.44E-08 0.32 221 4.26E-09 Eu-154 5.00E-11 95 3.51E-08 0.25 173 3.33E-09
% Change from the Base case = 100*(Sensitivity Case- Base Case)/Base Case Peak Radioactivity Radioactivity Sorbed Concentration in Solution Sorbed Concentration Nuclide pCi/L pCi pCi pCi/g H-3 0.0 0.0 Co-60 -16.2 -16.2 0.6 -16.2 Ni-63 -16.3 -16.3 0.5 -16.3 Sr-90 -15.8 -15.8 1.1 -15.8 Cs-134 -16.7 -16.7 -0.1 -16.7 Cs-137 -16.7 -16.7 0.0 -16.7 Eu-152 -17.5 -17.5 -1.0 -17.5 Eu-154 -16.6 -16.6 0.0 -16.6 36 Page 43 of 56
TSD 14-009 Revision 1 A.7: Low Density The density was decreased to 1.1 g/cm3 from the base case value of 1.5 g/cm3. Reducing the density caused an increase in both the solution concentration and the sorbed concentration. The increase was inversely proportional to the density. The change in density did not impact non-sorbing nuclides.
Diffusion Peak Radioactivity Radioactivity Sorbed Kd Coefficient Concentration in Solution Sorbed Concentration Nuclide (cm2/s) (ml/g) pCi/L pCi pCi pCi/g H-3 5.50E-07 0 8.70E-04 6267 0.0 0.00E+00 Co-60 4.10E-11 223 1.78E-08 0.13 125.8 3.97E-09 Ni-63 1.10E-09 62 1.44E-06 10.37 2829.8 8.93E-08 Sr-90 5.20E-10 2.3 1.31E-05 94.37 955.0 3.01E-08 Cs-134 3.00E-09 45 4.71E-07 3.39 671.8 2.12E-08 Cs-137 3.00E-09 45 1.79E-06 12.89 2553.1 8.06E-08 Eu-152 5.00E-11 96 7.25E-08 0.52 220.6 6.96E-09 Eu-154 5.00E-11 95 5.74E-08 0.41 172.8 5.45E-09
% Change from the Base case =
100*(Sensitivity Case- Base Case)/Base Case Peak Radioactivity Radioactivity Sorbed Concentration in Solution Sorbed Concentration Nuclide pCi/L pCi pCi pCi/g H-3 0.0 0.0 Co-60 36.9 36.9 0.4 36.9 Ni-63 37.1 37.1 0.6 37.1 Sr-90 33.1 33.1 -2.4 33.1 Cs-134 38.1 38.1 1.3 38.1 Cs-137 35.6 35.6 -0.6 35.6 Eu-152 34.8 34.8 -1.2 34.8 Eu-154 36.3 36.3 0.0 36.3 37 Page 44 of 56
TSD 14-009 Revision 1 Appendix B: Calculations to address Request for Additional Information (RAI) - 21.
The NRC Request for Additional Information 21 (RAI-21) is listed below.
Basis: In TSD 14-009, a sensitivity analysis was performed for a well receptor located outside of the basements. In this analysis, the peak concentrations in a receptor well 2 m outside the turbine basement were calculated. This analysis considered source terms from the auxiliary and turbine basements with contamination levels of 1 pCi/m2 and 0.001 pCi/m2 respectively. TSD 14-009 states that the well location was selected to be the closest place to put a well outside of the auxiliary basement, which is the basement that will have the highest levels of residual contamination. It is not clear if sources from other basements, such as the containment basement basements, could affect the groundwater at this location.
Also, it is not clear how the assumed concentrations will compare to the end state concentrations in those basements.
Summary Response:
Additional simulations were performed to examine contributions from the Containment to the Auxiliary basement through the Turbine basement to a receptor well. Two additional simulations were performed to determine if the Containment Basement impacts peak concentration in the receptor well. The first simulation assumed that the Auxiliary Basement and Containment Basement were contaminated to a fixed fraction of their Basement Inventory Limit (BIL). In this simulation, the Turbine basement contamination was scaled from its BIL based on observed concentrations of Cs-137. In the second simulation, the Containment contamination was set to zero. This simulation can be compared to the first to determine the impacts of Containment Basement on the receptor well concentration. The results showed that due to the transport time from the Containment Basement to the receptor well contamination in the Containment Basement had no impact on peak concentration in the Auxiliary Basement for less mobile nuclides (Co-60, Ni-63, Cs-134, Cs-137, Eu-152, and Eu-154). The key point of the analysis is that the Turbine Basement concentrations drive the peak concentration in the receptor well for short-lived less mobile nuclides (Co-60, Cs-134, Eu-152, and Eu-154) even if the inventory is several orders of magnitude lower than in the Auxiliary Basement. For two nuclides (Ni-63 and Cs-137) their longer half-life allows nuclides released from the Auxiliary Basement to reach the receptor well, but at levels that are 1 to 10% of their value in the Auxiliary Basement. For Ni-63 and Cs-137 nuclides released from the Containment Basement do not impact peak receptor well concentrations in these simulations. For mobile nuclides, e.g. H-3 and Sr-90, the impact of the Auxiliary Basement can be seen in the peak concentration. Releases from the Containment Basements at the same level as the Auxiliary Basement led to a slight increase (<20%) in the peak receptor well concentration for H-3 and Sr-90.
B1. Introduction Calculations were performed in the main body of this report to demonstrate the conservatism of the assumptions regarding placing a drinking water well within each basement. With current plans to backfill with clean concrete demolition debris (CCDD) it is likely that the pH will make any water in these basements non-potable. In addition, a well within a basement would likely be an unreliable water source due to recharge being from precipitation only. Therefore, an analysis was performed at the nearest downstream location, outside of the basements, that a well could be placed. At the Zion plant this is 2 m downstream from the Turbine Basement. Other basements are connected to 38 the Turbine Basement through penetrations that could lead to water from other basements Page 45 of 56
TSD 14-009 Revision 1 entering the Turbine Basement and ultimately reaching the well. These basements include the Auxiliary Basement which is directly connected to the Turbine Basement through penetrations and both reactor containment basements which have penetrations that connect to the Auxiliary Basement.
Characterization data and demolition plans suggest that the majority of the residual contamination will be in the Auxiliary Basement at the time of license termination. Based on characterization data, the Turbine Basement currently contains minor amounts of contamination now, before any remediation is performed. The Containment Basements will have all activated and contaminated concrete removed during demolition leaving only the underlying steel liner which is expected contain very low levels of residual radioactivity at the time of license termination. However, the levels of residual radioactivity remaining on the liner will not be available until after the concrete is removed.
The NRC requested additional information regarding the original evaluation of a well located outside of the basements. Specifically, whether residual radioactivity in the Containment Basements could be transported through the Auxiliary Basement and through the Turbine Basement to the receptor well, and if so what the impact of this transport would be on the estimated concentrations in the well.
To accommodate the NRC request a few changes to the original receptor well sensitivity model were required. The major changes from the main body of the report is the addition of the Containment Basement and changing the source term from 1 pCi/m2 of surface area in the basement to amounts proportional to the Basement Inventory Limit (BIL) for the Containment and Auxiliary Basements. This approach is conservative in that it maximizes the potential source terms for Containment and Auxiliary basements. Note that it is highly unlikely that the residual radioactivity remaining in the Containment basement at the time of license termination would approach the BIL. The Turbine basement Inventory is based on current radiological conditions which were conservatively estimated from characterization data. For consistency with the Containment and Auxiliary source terms the Turbine inventory is expressed as a fraction of the Turbine basement BIL. Simulating all three basements simultaneously also led to changes in the simulated geometry. The changes resulting from this will be discussed as part of the model development and source term descriptions that follow.
B2. Conceptual Models of Release B2.1 Site Overview Major features at the Zion Nuclear Power Station include two reactor Containment Basements (Unit-1 and Unit-2 in Figure 1, a Fuel Handling Basement, Auxiliary Basement, Turbine Basement, Crib House, and Waste Water Treatment Facility (WWTF).
The proposed decommissioning approach involves removal of regions with high-levels of contamination through a remediation process. There will be some surface contamination and volumetric contamination left in place. This contamination will provide a potential source of radioactivity to the groundwater. These structures will be filled with non-contaminated material.
Currently CCDDis the likely fill material.
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TSD 14-009 Revision 1 There are seven basements that will have residual structures beginning three feet below grade. For the purposes of this analysis consideration is given to the Containment Basements, Auxiliary Basement and the Turbine Basement which are the primary potential contributors to a downstream receptor well. The Fuel Handling basement is expected to contain residual radioactivity but adding this structure to the model will not affect the result because of the small relative volume of the Fuel Handling basement but more importantly, because the inventory in the Containment basement is assumed to be at the BIL and water from the Fuel handling basement would flow through Containment before reaching the Auxiliary basement. The maximum groundwater concentrations in each basement are by definition equal. Adding a source with the same concentrations would not increase the concentrations in the water flowing from Containment to the Auxiliary basement. The WWTF and Crib House would not contribute to the receptor well due to their locations.
B2.2 Modeling Overview The Disposal Unit Source Term - Multiple Species (DUST-MS) computer code has been selected to calculate the source term release and equilibrium water concentration at the receptor well which is assumed to be in the center of the backfilled basement. DUST-MS has received wide-spread use in subsurface radionuclide release calculations and undergone model validation studies (Sullivan, 1993; 2006). To maintain consistency between all calculations DUST-MS was used for all simulations. The nuclides of concern and model parameters are provided in the main body of the report.
B2.3 Release Models For the instant release model the key parameters are the distribution coefficient (Kd), porosity and bulk density of the fill material. For the conceptual model using a receptor well downstream from the Turbine Basement, the Turbine and Containment Basements were modeled using instant release. For consistency with the base case, the Auxiliary Basement was modeled with diffusion controlled release to simulate the release out of the contaminated concrete. The instant release model is the most conservative approach as the entire inventory is available immediately for transport.
B2.4 Model Geometry DUST-MS is a one dimensional model. The conceptual model contains contaminated surfaces in the direction of flow. DUST-MS model requires a flow area to calculate the correct concentrations above the floor. The flow area is defined as the area perpendicular to the transport direction. In these simulations, the transport direction is towards the Lake. Therefore, the flow is the product of the height of the water table above the floor and the width of the basement that is parallel to the Lake. Table 4 provides the height to the water table based on a 579-foot elevation, effective distance parallel to the Lake, flow area, and effective length of the contaminated zone. The product of the flow area and length of the contaminated zone gives the total volume for each basement.
These widths, height to the water table, and volumes were calculated by ZionSolutions staff (Farr, 2014) and are provided in Table 1.
Table B-1 Actual Basement Geometry.
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TSD 14-009 Revision 1 Width Height Flow or to Void Area Contaminated Radius Water Space to Structure (m2) Zone Length m Table WT (m) m m3 Containment Basements 20.95 4.27 140.4 44.8 6537 Auxiliary Basement 80.11 11.28 903 31.5 28445 Turbine Basement 145.1 5.79 571.5 29 26135 Simulation of flow from the Containment Basements to the Auxiliary Basement and then to the Turbine Basement is not possible within the constraints of a one dimensional model without assumptions. In this case, the Containment Basement is placed directly behind the Auxiliary Basement. The Auxiliary Basement is modeled as being 31.5 m in the direction of flow (Table 4).
The Containment Geometry in Table 4 will be modified to account for the differences in Flow Area between the Containment and Modeled Geometry.
The modeled geometry is presented in Figure 1. The width of the Auxiliary Basement is 80.1 m, which is less than the Turbine Basement. The one-dimensional simulation requires that the width perpendicular to flow remain constant. Therefore, for this simulation only the portions of the Turbine Basement downstream from the Auxiliary Basement are modeled. The length of the Turbine Basement parallel to flow is 29.3 m. Therefore, the total volume of the Turbine Basement in this simulation is reduced as compared to the actual volume. This is addressed in the next section. The receptor well is 2 meters downstream of the Turbine Basement.
The one-dimensional simulation also requires the depth to the water table to remain the same in all basements. The actual depth to the water table is deeper in the Auxiliary Basement as compared to the Turbine Basement. The geometry and flow direction requires that any release from the Auxiliary Basement travel through the Turbine Basement. Therefore, the appropriate depth to the water table for this simulation is that of the Turbine Basement, 5.79 m (19 ft.). Reducing the height of the water table in the Auxiliary Basements reduces the simulated volume of the Auxiliary Basement by the ratio of the water table levels (5.79 m/11.28 m). The total area available for flow (basement width multiplied by the height to the water table) is 463.7 m2. The modeled geometry is presented in Table 2. The adjustment in volumes maintains the concentrations consistent with the mixing bath model and therefore, is needed for consistency with the BIL levels and does not impact the conclusions of the receptor well model.
Table B-2 Modeled Geometry Width Height Flow or to Void Area Contaminated Radius Water Space to Structure (m2) Zone Length m Table WT (m) m m3 Containment Basements 80.11 5.79 464 14.1 6537 Auxiliary Basement 80.11 5.79 464 31.5 14616 Turbine Basement 80.11 5.79 41464 29 13456 Page 48 of 56
TSD 14-009 Revision 1 B-2.4.1 Receptor Well The Auxiliary Basement will have the highest levels of residual contamination. The Auxiliary Basement is adjacent to the Turbine Basement and there are penetrations that will remain in place and connect these basements. The Containment Basements are also connected to the Auxiliary Basement by penetrations and this is also modeled by adding a third contaminated zone to represent the Containment Basement.
The closest place to put a well in the shallow aquifer outside of the Auxiliary Basement is just outside and to the east of the Turbine Basement. The Containment Basements and the Auxiliary Basement foundations rest on the clay aquitard and a well located directly to the east of the Auxiliary basement, and under the Turbine Basement floor would not flow. To examine the maximum concentration that could be obtained from a well in the soil, DUST-MS was used to predict the concentrations 2 meters outside of the eastern edge of the Turbine Basement, Figure 3.
Therefore, the modeled domain contains the three contaminated regions representing the Containment Basement, Auxiliary Basement and the section of the Turbine Basement that aligns with the Auxiliary Basement and groundwater flow direction, Figure 1. A schematic representation of the model domain is presented in Figure 1. The dotted rectangular region is the modeled region and consists of clean soil upstream from the Auxiliary Basement, the Auxiliary and Turbine Basements and clean soil downstream of the Turbine Basement. A hypothetical well located 2 m from the edge of the Turbine Basement is shown.
Groundwater Containment Flow Basement Direction Auxiliary Basement Turbine Basement Well Soil Figure B-1 Schematic Representation of Flow the geometry used to assess flow to a well 42 outside the Turbine Basement.
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TSD 14-009 Revision 1 B3. Analysis Parameters All analysis parameters for material properties are presented in the main body of this report. This section will highlight any differences between those values and values in the sensitivity study.
B3.1 Parameters Initial conditions assumed that the groundwater concentration of each contaminant was zero everywhere. The source term is discussed in the following section.
The distribution coefficients (Kd) are important parameters in controlling the equilibrium concentrations and transport. They remain the same as in the base case study, but are repeated here because they have a direct impact on the inventory needed to reach 1 pCi/L. The greater the Kd value, the greater the inventory to reach this fixed level in the water concentration. Selected values are in Table 3.
The other key parameters that influence the initial inventory are the bulk density 1.5 g/cm 3 and the porosity (0.25). These values are unchanged from the base case. For the diffusion release in the Auxiliary Basement the diffusion coefficients used in the base case were used and they are presented in Table 4.
Table B-3 Selected distribution coefficients (Sullivan, 2014)
Radionuclide Half Life (years) Basement Fill Kd to Be Used cm3/g H-3 12.3 0 Co-60 5.27 223 Ni-63 96 62 Sr-90 29.1 2.3 Cs-134 2.06 45 Cs-137 30 45 Eu-152 13.4 95 Eu-154 8.2 95 43 Page 50 of 56
TSD 14-009 Revision 1 Table B-4 Typical diffusion coefficients in cement for radionuclides of concern Nuclide Diffusion Coefficient Selected Diffusion Range (cm2/s) Coefficient (cm2/s)
H-3 6.0E 5.5E-07 5.5E-07 Co-60 5.0E 4.1E-11 4.1E-11 Ni-63 8.7E 1.1E-09 1.1E-09 Sr-90 1.0E 5.2E-10 5.2E-10 Cs-134; Cs-137 4.0E 3.0E-09 3.0E-09 Eu-152; Eu-154 1.0E 5.0E-11 5.0E-11 B-3.1.1 Source Term Inventory To bound the release, the inventories in the Containment and Auxiliary basements are scaled to the BIL. The BIL is the inventory that would lead to a dose of 25 mrem/y based on all pathways using the Basement Fill Model. The Turbine Basement uses the BIL scaled to a conservative estimate of the Cs-137 inventory remaining based on characterization.
The initial analysis presented in TSD 14-009 Revision 0 assumed that the activity in the Turbine basement was lower than the Auxiliary basement by a nominal factor of 1E-03. The 1E-03 value was the estimated fraction of the actual inventory in the Turbine basement based on characterization data to the Turbine Basement Inventory Limit (BIL), i.e., the activity that would result in 25 mrem/yr. The addition of the Containment basement required the method for calculating the source term for the well receptor sensitivity analysis to be slightly revised. This included a more accurate calculation of the Turbine inventory and BIL fractions.
From ZSRP LTP Chapter 2, Table 2-25 the average concentration from biased cores from the Turbine basement floor was 15.4 pCi/g Cs-137. This activity was found in sporadic localized areas but for the purposes of this sensitivity analysis the 15.4 pCi/g concentration was conservatively assumed to be uniformly distributed over the entire basement floor. From TSD 14-021, Table 2, the Turbine basement floor surface area is 48576 ft2 or 4512 m2. The total mCi, assuming 15.4 pCi/g uniformly distributed over the entire floor area, at a depth of 0.5 inch, with a concrete density of 2.4 g/cm3 was determined to be 2.12 mCi.
ZSRP LTP Chapter 6, Table 6-16, lists the BFM Groundwater Dose Factors. Since the well sensitivity analysis addresses the groundwater scenario only (not Drilling Spoils) the Groundwater Dose Factors are used as opposed to the full Basement Dose Factors in LTP Chapter 6, Table 6-18.
From ZSRP LTP Chapter 6 Table 6-16, the BFM Groundwater Dose Factor for Cs-137 in the Turbine basement is 3.92E-02 mrem/yr per mCi. Therefore, the Cs-137 BIL for the groundwater scenario is calculated as (1/3.92E-02 mrem/yr per mCi) x 25 mrem/yr = 637 mCi. Given that the actual inventory of Cs-137 in the Turbine basement is 2.12 mCi as shown above, the ratio of actual inventory to the BIL is 2.12/637 = 3.32E-03.This value was approximated as 1E-03 in the original well receptor sensitivity analysis.
The Turbine BILs for all ROCs were multiplied by the 3.32E-03 fraction to determine the Adjusted Turbine inventory which was used in the well sensitivity analysis. The inventory for all ROC in the Containment and Auxiliary basements were conservatively assumed to be at the maximum allowable level, i.e., the BIL. The resulting total inventories for each ROC in each basement applied to the well sensitivity analysis are provided in Table 5. Note that both the BIL and final adjusted inventories are listed for the Turbine basement.
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TSD 14-009 Revision 1 Table B-5 Inventory based on BIL and Turbine Basement Characterization Data Adjusted Auxiliary Containment Turbine Turbine Radionuclide mCi mCi mCi mCi H-3 4.03E+03 9.20E+02 3.68E+03 1.22E+01 Co-60 2.50E+05 2.19E+03 8.72E+03 2.90E+01 Ni-63 8.75E+04 1.56E+04 6.23E+04 2.07E+02 Sr-90 7.59E+01 5.54E+00 2.21E+01 7.34E-02 Cs-134 2.70E+03 1.27E+02 5.06E+02 1.68E+00 Cs-137 9.48E+02 1.59E+02 6.37E+02 2.12E+00 Eu-152 4.20E+05 6.46E+03 2.58E+04 8.57E+01 Eu-154 3.69E+05 4.45E+03 1.78E+04 5.90E+01 The Base Case used a uniform contamination level of 1 pCi/m2 in each basement. This lead to an inventory of 6503 pCi in the Auxiliary Basement, 2759 pCi, and 14679 pCi for each nuclide. For comparison with the Base Case, the units of inventory were scaled to pCi in the following analysis.
The objective of this sensitivity study was to show the relative contribution at the receptor well for each basement. Scaling does not impact this objective.
The BIL values are based on the entire volume of the structure that will remain at Zion. Therefore, the inventories of the Auxiliary basement and Turbine basement have to be reduced to match the modeled geometry. This reduction is the ratio of the modeled volume, Table 2, to the actual volume, Table 1. This reduces the Auxiliary Basement inventory by a factor of 0.51 and the Turbine Basement by a factor of 0.55. The resulting inventory used in the simulation is presented in Table 6.
Table B-6 Inventory used in the simulation.
Nuclide Auxiliary Containment Turbine pCi pCi pCi H-3 2.07E+03 9.20E+02 6.75E+00 Co-60 1.28E+05 2.19E+03 1.60E+01 Ni-63 4.49E+04 1.56E+04 1.14E+02 Sr-90 3.90E+01 5.54E+00 4.05E-02 Cs-134 1.38E+03 1.27E+02 9.28E-01 Cs-137 4.87E+02 1.59E+02 1.17E+00 Eu-152 2.15E+05 6.46E+03 4.73E+01 Eu-154 1.89E+05 4.45E+03 3.26E+01 The source term used for the well receptor sensitivity analysis is provided in Table 7. The pCi/m 2 units are used for consistency with the base case but as stated above the results of the sensitivity analysis is relative and the units do not impact the results. Table 7 is generated by dividing the values in Table 6 by the surface area in each basement.
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TSD 14-009 Revision 1 Table B-7 Inventory per unit surface area in each basement.
Nuclide Auxiliary Containment Turbine pCi/m2 pCi/m2 pCi/m2 H-3 6.19E-01 3.34E-01 8.32E-04 Co-60 3.84E+01 7.92E-01 1.97E-03 Ni-63 1.35E+01 5.64E+00 1.41E-02 Sr-90 1.17E-02 2.01E-03 5.00E-06 Cs-134 4.15E-01 4.59E-02 1.15E-04 Cs-137 1.46E-01 5.78E-02 1.44E-04 Eu-152 6.45E+01 2.34E+00 5.84E-03 Eu-154 5.67E+01 1.61E+00 4.02E-03 B.3.1.2 Receptor Well Parameters for Transport Model Receptor well transport parameters are identical to those used in the sensitivity analysis in the main body of the report. They are repeated here in Table 8.
Table B-8 Transport Parameters used to calculate peak concentrations in a receptor well located outside of the basements.
Parameter Value Reference Soil Density 1.81 (g/cm3) CRA, 2014 Soil Effective Porosity 0.29 CRA, 2014 Groundwater Darcy Velocity 41.6 m/y CRA, 2014 3
Soil Kd: Co-60 223 (cm /g) Sullivan, 3
Ni-63 62 (cm /g) 2014 3
Sr-90 2.3 (cm /g)
Cs-134 45 (cm3/g)
Cs-137 45 (cm3/g)
Eu-152 95 (cm3/g)
Eu-154 95 (cm3/g)
In the simulation, the Auxiliary Basement is modeled with 100 cells giving a total length parallel to flow of 31.5 m. The Turbine Basement is modeled with 92 cells for a total length of 29 m. The Containment Basements are modeled with 45 cells for a total length of 14.1 m. The adjustment in length for the Containment Basement is needed to match the flow area used in the simulation with the total Containment volume.
Three modeling scenarios considered are:
a) High contamination in the Auxiliary and Containment Basements (scaled BIL for all contaminants) and lower levels of contamination in the Containment and Turbine Basements (BIL scaled to match characterization data).
b) High contamination in the Auxiliary Basement (scaled BIL for all contaminants), a lower level in the Turbine Basement (BIL scaled to match characterization data), and zero contamination in 46 Page 53 of 56
TSD 14-009 Revision 1 the Containment Basement. The results of this will be used to determine the impacts of the Containment Basement on the receptor well concentration.
B4 Results B4.1 Base Case The base case consists of contamination is the inventory levels in Table 6. Table 9 provides the predicted peak concentrations in the center of the Containment, Auxiliary and Turbine Basements, and at the Receptor Well located 2 meters outside the Turbine Basement. The ratio of the concentration at the well to that in the Auxiliary Basement is a measure of the impacts of transport on concentration and reflects the conservatism in assuming a well is located in the center of the Auxiliary Basement.
Table B-9 Predicted peak concentrations for the base case.
Ratio Well to Auxiliary Time Receptor Basemen to Containment Auxiliary Turbine Well t peak (pCi/L) (pCi/L) (pCi/L) (pCi/L) years H-3 5.6E-04 5.6E-04 4.2E-04 3.9E-04 0.70 1.1 Co-60 9.93E-07 9.95E-07 3.5E-09 1.7E-14 1.7E-08 16.8 Ni-63 2.54E-05 2.57E-05 3.3E-06 1.9E-06 0.07 286 Sr-90 2.3E-07 2.4E-07 1.7E-07 1.6E-07 0.65 13.6 Cs-134 2.83E-07 2.85E-07 1.0E-09 1.7E-13 0.00 5 Cs-137 3.6E-07 3.6E-07 4.9E-09 1.8E-09 0.01 167 Eu-152 6.9E-06 6.9E-06 2.5E-08 2.5E-10 0.00 21.2 Eu-154 4.7E-06 4.7E-06 1.7E-08 3.8E-11 0.00 15.3 A few key points can be determined from Table 9:
For mobile nuclides (H-3 and Sr-90) the dilution from the Auxiliary Basement to the well is less than a factor of 2. This reflects the fast transport time.
The Containment and Auxiliary Basement have the same peak concentration. This is expected as the inventory in both was set to the same fraction of the BIL. The BIL is the inventory that will give a dose of 25 mrem/yr.
For short lived and less mobile nuclides the time to peak is less than 20 years. This suggests that contamination from the Auxiliary Basement the major contribution to the receptor well and that Auxiliary Basement contributions to the well are minor. For these nuclides, the peak concentrations are controlled by the contamination in the Turbine Basement.
For Ni-63 and Cs-137 they do not reach their peak concentration for several hundred years suggesting the higher contamination in the Auxiliary Basement reaches the well. However, the travel time allows for substantial decay and only peak concentrations in the well are 1%
for Cs-137 and 10% for Ni-63 of their value in the Auxiliary Basement.
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TSD 14-009 Revision 1 B4.2 Zero Contamination Levels in the Containment In this scenario the contamination level in the Containment Basement was set to zero. This permits an evaluation of the impact of Containment on the receptor well concentrations. Peak concentrations, the ratio of the peak value in the receptor well to that in the center of the Auxiliary Basement and the time to reach the peak in the well are provided in Table 10.
Table B-10 Predicted peak concentrations when the Containment has no contamination.
Ratio Well to Auxiliary Time Turbine Receptor Basemen to Containment Aux Bldg Bldg Well t peak (pCi/L) (pCi/L) (pCi/L) (pCi/L) years H-3 0 5.4E-04 3.4E-04 3.2E-04 0.59 1 Co-60 0 1.0E-06 3.5E-09 1.7E-14 1.7E-08 16.8 Ni-63 0 2.6E-05 3.3E-06 1.9E-06 0.07 286 Sr-90 0 1.7E-07 1.3E-07 1.2E-07 0.70 13.6 Cs-134 0 2.9E-07 1.0E-09 1.7E-13 0.00 5 Cs-137 0 3.6E-07 4.9E-09 1.8E-09 0.01 167 Eu-152 0 6.9E-06 2.5E-08 2.5E-10 0.00 21.2 Eu-154 0 4.7E-06 1.7E-08 3.8E-11 0.00 15.3 The results from this scenario closely match the Base Case showing only a minor influence of the Containment Basements. For all nuclides with the exception of H-3 and Sr-90, removing all of the contamination from the Containment Basement did not lead to a change in receptor well concentration. For the mobile nuclides (H-3 and Sr-90) the peak well concentrations increase due to the contributions from the Containment Basement. The peak H-3 receptor well concentration increase 20% and the peak Sr-90 concentration increases 30%.
B5 Conclusions Two scenarios have been simulated to examine the potential of contaminated water from the Containment Basements could travel through penetrations to the Auxiliary Basement through another set of penetrations to the Turbine Basement and to a receptor well 2 m down gradient of the Turbine Basement. The Base Case assumed contamination level based on the BIL for each nuclide in the Containment and Auxiliary Basement and contamination in the Turbine Basements ratioed to the measured value of Cs-137. The key findings are:
For mobile nuclides (H-3 and Sr-90) it does not make much difference where the contamination is located as it will reach the receptor well. Adding the contribution from the Containment at the same level as in the Auxiliary Basement led only to a 20 - 33%
increase in concentration at the well.
For Ni-63 and Cs-137 due to their long half-lifes and somewhat low distribution coefficients (62 for Ni-63 and 45 for Cs-137 in this simulation) contamination in the Containment Basement and Auxiliary Basement will reach the receptor well, but at peak concentrations less than 10% (Ni-63) and 1% (Cs-137) of their value in these basements.
For less mobile nuclides with short half-life (Co-60, Cs-134, Eu-152 and Eu-154) very little contamination (<0.1%) in the Auxiliary Basement will reach the receptor well.
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TSD 14-009 Revision 1 Having contamination in the Containment Basement at the same level as in the Auxiliary Basement had no impact on peak receptor well concentrations for all nuclides except H-3 and Sr-90. These two nuclides moved fast enough to prevent substantial decay before reaching the receptor well.
B6 References Conestoga-Rovers & Associates, Evaluation of Hydrological 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.
Milian, L., T. Sullivan. Sorption (Kd) measurements on Cinder Block and Grout in Support of Dose Assessments for Zion Nuclear Station Decommissioning, Brookhaven National Laboratory Report to ZionSolutions, June 24, 2014, BNL-1055441-2014-IR.
Sullivan, T.M., "DUST - Disposal Unit Source Term: Data Input Guide." NUREG/CR-6041, BNL-NUREG-52375, 1993.
Sullivan, T.M., "DUSTMS_D - Disposal Unit Source Term - Multiple Species - Distributed Failure Data Input Guide. Rev 1., BNL-75554-2006, Brookhaven National Laboratory, Upton, NY, 11973, January, 2006.
Sullivan, T.M., Recommended Values for the Distribution Coefficient (Kd) to be Used in Dose Assessments for Decommissioning the Zion Nuclear Power Plant, BNL 105542-2014, June, 9, 2014.
Sullivan, T.M., Evaluation of Maximum Radionuclide Groundwater Concentrations for Radionuclides of Concern Zion Station Restoration Project, Brookhaven National Laboratory, Draft Letter Report, December 3, 2014.
Yim, S.P, T.M. Sullivan, and L. Milian, Sorption (Kd) measurements in Support of Dose Assessments for Zion Nuclear Station Decommissioning, Brookhaven National Laboratory Report to ZionSolutions, December 12, 2012, BNL-105981-2012-IR.
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