ML16211A374

TSD-14-009 Rev. 2, Brookhaven National Laboratory: Evaluation of Maximum Radionuclide Groundwater Concentrations for Basement Fill Model.

ML16211A374
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Site:	Zion  File:ZionSolutions icon.png
Issue date:	07/20/2016
From:	ZionSolutions
To:	Office of Nuclear Material Safety and Safeguards
References
ZS-2016-0084 TSD 14-009, Rev. 2
Download: ML16211A374 (43)

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~ :::::===:1 ZionSolutions, LLC. ZIONSOLUTIONSuc Technical Support Document TSD 14-009 Brookhaven National Laboratory: Evaluation of Maximum Radionuclide Groundwater Concentrations for Basement Fill Model Revision 2 Originator: Terry Sullivan Date: 6/1/2016 Brookhayen National Laboratory Reviewer: -------------------------- Date: 6/1/2016 David Fauver Approval: __ ---=:.~ ~----~_+c_~.

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TSD 14-009 Revision 2 BNL-107250-2016-IR-R2 Evaluation of Maximum Radionuclide Groundwater Concentrations for Basement Fill Model Zion Station Restoration Project Terry Sullivan Brookhaven National Laboratory Revision 2 June 1, 2016 Page 2 of 43

TSD 14-009 Revision 2 Revision Log Section Page Rev. Date Reason(s) for Revision Title page 2 5/20/16 Revised Revision number Title page 2 5/20/16 Revised effective date Title page 17, 2 5/20/16 Changed the total pCi released for H-3 in Table Table 10 from 6503 pCi to 6467 pCi. The value 6503 10 pCi is the total inventory in the analysis and this can not be the total released. This simulation has diffusion controlled release from the concrete and some H-3 decays in the concrete prior to release.

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TSD 14-009 Revision 2 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 Page 4 of 43

TSD 14-009 Revision 2 Appendix A: Sensitivity Analysis Results ................................................................................... 31 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 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 5 of 43

TSD 14-009 Revision 2 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 6 of 43

TSD 14-009 Revision 2 Evaluation of Maximum Radionuclide Groundwater Concentrations for Radionuclides of Concern Zion Station Restoration Project

1. Introduction ZionSolutions is i n t he pr ocess of d ecommissioning t he Zion N uclear P ower Station ( ZNPS).

After d ecommissioning is c ompleted, t he s ite will contain t wo r eactor Containment Buildings, the Fuel Handling Building and T ransfer C anals, Auxiliary Building, Turbine Building, C rib 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 d iameter i ntake and d ischarge pipes. Th ese additional structures are n ot included i n t he m odeling d escribed i n t his r eport but t he i nventory remaining ( expected t o b e very low) will be included with one of the structures that are modeled as designated in the Zion Station R estoration P roject ( ZSRP) License T ermination P lan ( LTP). The r emaining underground structures will be backfilled with clean material. The final selection of fill material has not been made.

Remaining structures will contain r esidual radioactive m aterial to va rying e xtents. The bul k o f 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 co ncrete ch aracterization d ata, t he h ighest en d s tate s ource t erm i s an ticipated t o b e contained in the Auxiliary Building floor located approximately 50 feet below grade. The end state s ource te rm w ill b e at l east 3 f eet below grade i n all remaining structures eliminating conventional pa thways s uch a s di rect r adiation a nd i nhalation r endering g roundwater r elated pathways the most significant potential sources of future exposure.

An i mportant c omponent of t he de commissioning pr ocess i s t he de monstration t hat a ny remaining a ctivity w ill not c ause a h ypothetical in dividual (average m ember o f t he c ritical group) to receive a dose in excess of 25 mrem/y as specified in 10 CFR Part 20 Subpart E. To demonstrate compliance w ith 10 CFR Part 20 S ubpart E r equires m odeling o f t he fate and transport of r adioactive material t o a r eceptor. This i nvolves c haracterization of t he bui lding basements t o r emain on s ite t o qua ntify t he a mount of r esidual r adioactivity, modeling th e release o f r adioactivity from t he co ncrete, and mixing w ith th e w ater c ontained in th e f ill material. Transport away from the fill to a r eceptor w ell located outside of t he 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 c ontain th e ma jority o f th e r esidual c ontamination inventory a t th e time o f lic ense termination. This analysis was used b y ZSRP along with characterization data and RESRAD modeling t o s creen out l ow dos e s ignificance r adionuclides a nd i dentify eight radionuclides of concern (ROCs) Co-60, N i-63, S r-90, Cs-134, C s-137, Eu -152, E u-154, a nd H -3 for d etailed assessment.

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TSD 14-009 Revision 2 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 t he m iddle of t he s ubsurface r emains of t he s even bui ldings. T he r atio of t he r esulting equilibrium water concentration in units of picocuries per liter (pCi/L) to the assumed inventory in units of C uries (Ci) f or e ach bui lding i s us ed b y ZSRP, i n c onjunction w ith t he R ESRAD code, to demonstrate compliance with 10 CFR 20 Subpart E.

Calculation of t he fill in terstitial w ater c oncentration r equires s ite-specific i nformation on t he hydrogeologic properties (effective porosity and bulk density) and chemical transport properties (sorption). C onestoga-Rovers & A ssociates ( CRA) h as co llected a s ubstantial am ount o f s ite-specific hydrogeologic data (CRA, 2014).

Brookhaven N ational Laboratory (BNL) has determined s ite-specific sorption da ta f or fi ve nuclides that a re R OCs w ith four s oil t ypes, two concrete types o f construction de molition 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 r eport ( Sullivan, 2014) pr ovided r ecommended va lues t o us e f or dos e assessment based on measured values, when available, and literature values in other cases. F or 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 pr esent a simplified conceptual m odel f or release f rom t he buildings w ith r esidual subsurface s tructures that can be us ed t o pr ovide a n uppe r bound on c ontaminant concentrations in the fill material.

b) Provide maximum water concentrations and the corresponding amount of mass sorbed to the s olid fill material that c ould oc cur i n e ach bui lding for use i n d ose a ssessment 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 2

2. Conceptual Models of Release 2.1 Site Overview Figure 1 provides t he s ite l ayout a t ZNPS located on t he s hores of Lake M ichigan. Major features i nclude two r eactor Containment Buildings ( Unit-1 a nd U nit-2 i n Figure 1, a F uel Handling Building, Auxiliary Building, Turbine Building, C rib H ouse, a nd W aste W ater Treatment Facility (WWTF).

The pr oposed de commissioning a pproach involves removal of r egions w ith h igh-levels o f contamination t hrough a r emediation pr ocess. There will be some surface c ontamination a nd volumetric c ontamination le ft in p lace. T his c ontamination w ill p rovide a p otential s ource o f 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. T he fill may contain a combination of t he t hree r emaining c hoices or it c ould onl y i nclude s and. C inder bl ock 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. C ontaminated c oncrete f rom i nside t he l iner i n t he C ontainment B uildings will be removed and t his w ill s ubstantially d ecrease t he a mount of c ontamination i n t he C ontainment 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 U nit S ource Term - Multiple S pecies (DUST-MS) computer co de h as b een selected to calculate the source term release and equilibrium water concentration at the receptor well w hich i s assumed t o be i n t he c enter o f t he ba ckfilled bui lding. D UST-MS h as r eceived wide-spread use in subsurface radionuclide release calculations and undergone model validation studies ( Sullivan, 1993; 2006) . T he equilibrium m odel c an be e asily c alculated b y hand.

However, DUST-MS is necessary when simulating diffusion controlled release or transport to a receptor w ell. T o m aintain co nsistency b etween al l cal culations D UST-MS w as u sed f or al l simulations.

An imp ortant p arameter is th e volume of w ater av ailable t o mix w ith released r adionuclides.

Another important parameter defines how the release of contaminants will be modeled. In many 3

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TSD 14-009 Revision 2 buildings the contamination is expected to be loosely bound or near the surface of the remaining structure. In t hese bui ldings, t he r elease i s a ssumed t o oc cur i nstantly, s uch t hat t he e ntire inventory i s a vailable immediately after lic ense te rmination. In s ome bui ldings t he contamination i s e xpected t o have diffused into t he c oncrete resulting in v olumetrically contaminated co ncrete. For t hese bui ldings, a di ffusion controlled release m odel i s us ed. The Auxiliary Building has been ch aracterized 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 ex pected t o b e v olumetrically co ntaminated b elow t he l iner b ut t he e xtent of t his 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 s ummarizes th e to tal fill volume available f or mix ing and the release as sumptions f or 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. T his 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 2 Turbine I

,I 1;;:. -' -, 4J, D _._ I~~

Figure 1 Zion Site building layout.

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TSD 14-009 Revision 2 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 t he C ontainment B uildings onl y l oose s urface c ontamination i s e xpected t o r emain. The distribution of t he s urface s ource t erm i s generally expected t o be uni form ove r t he 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. T he minimal c ontamination p resent is assumed t o be on t he concrete surfaces an d instantly released.

Diffusion Controlled R elease was us ed t o estimate t he s ource t erm r elease r ate for t he Fuel Handling Building Basement and Fuel Transfer Canals due to expected contamination at depth in concrete after the liners are removed.

In a ddition to th e p rimary mo deling used for 10 C FR 20 S ubpart 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 2 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. O utside 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 pC i/m2 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 f or s caling w ith t he f inal i nventory m easured i n e ach b asement 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 r emediation is c ompleted, th e r esults o f th e s imulations p resented in th is report should be multiplied by the ratio of the measured inventory to simulated inventory.

Material p roperties w ere c hosen to ma tch s ite-specific va lues t o t he extent pos sible. S orption coefficient, K d, values were based on the measured values for Zion soils, concrete, cinder block, and grout (Yim, 2012, Milian, 2014) w hen a vailable a nd l iterature va lues w hen s ite-specific values were not available. A review of literature values and rationale for selecting K d for dose assessment w as p erformed ( Sullivan, 2014) . The K d values selected f rom t he l iterature w ere chosen t o give a co nservative es timate o f water concentration ( highest va lue) f or dos e assessment. W hen si te-specific v alues ar e av ailable, t he l owest K d value m easured i n an y f ill material or soil was selected.

The compliance a ssessment r equires pr ediction of t he r elease a nd t ransport of c ontaminants t o the h ypothetical i ndividual. C haracterization s tudies a nd a ssessments 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 2 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 (Ci)

Area (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 2 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 t he i nstant r elease model t he k ey p arameters ar e the d istribution c oefficient ( Kd), por osity and bulk density of the fill material. T he 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 t he buildings, A uxiliary and F uel, there is v olumetric contamination in th e c oncrete 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 ha ve be en conducted f or t he di ffusion i n c oncrete o f t he r adionuclides und er consideration a t Zion ( H-3, Co-60, N i-63, S r-90, C s-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 an d t he aggregate. A t ypical range from th e lite rature is p resented in T able 4. Th e 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 2 In t he conceptual m odel f or di ffusion c ontrolled r elease i t i s a ssumed t hat t he c oncrete is uniformly contaminated over a 0.5 inch thickness and t hat al l o f the m aterial i s r eleased at t he surface ( i.e. i t doe s not di ffuse f urther i nto t he c oncrete). T his a ssumption i s e quivalent t o having on e s ide o f t he contaminated z one a s a no f low bounda ry. In pr actice, s ome of t he nuclides would continue to diffuse deeper into the concrete initially and thereby increase the time before being released to the water. T he 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 s lab i s m odeled a s be ing one inch t hick and a llowed t o f low out of bot h s ides of t he s lab.

Using the principle of symmetry, the centerline is a no f low boundary and this is equivalent to having a s lab 0.5 inch t hick but pr eventing di ffusion further in to th e c ement. 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 r elease f rom o ne s urface o f t he 0.5 inch s lab. Using s ymmetry, t he r elease f rom 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 r ise to le vels th at make i t non pot able. Notwithstanding t he hi gh pH condition, t he conceptual m odel as sumes t hat t his w ater w ill b e u sed as a r esidential water s upply, l ivestock water supply and for irrigation. This section addresses a m ore credible scenario where the well is located outside of the basements.

The A uxiliary Building will h ave th e h ighest le vels o f residual contamination. T he A uxiliary Building is adjacent to the Turbine Building and there are penetrations that will remain in place and c onnect t hese bui ldings. T he C ontainment Buildings are also c onnected t o t he Auxiliary Building by penetrations but Containment will have minimal contamination after removal of all internal concrete.

The cl osest p lace t o p ut a w ell in t he s hallow a quifer outside o f th e A uxiliary B uilding is ju st outside and to the east of the Turbine Building. The Auxiliary Building foundation rests on the clay aquitard a nd a w ell l ocated di rectly t o t he east of t he A uxiliary bu ilding, a nd unde r t he Turbine B uilding f loor w ould not f low. To examine the m aximum c oncentration t hat c ould be obtained f rom a w ell in th e s oil, DUS T-MS w as us ed t o pr edict t he c oncentrations 2 m eters outside of the eastern ed ge of the Turbine Building, Figure 1. Therefore, the modeled domain contains th e A uxiliary Building and t he s ection of t he T urbine B uilding t hat a ligns w ith t he Auxiliary B uilding a nd groundwater f low di rection. A s chematic representation of t he m odel domain i s pr esented i n Figure 3. T he dot ted rectangular r egion i s t he m odeled r egion and consists of clean soil upstream from the Auxiliary Building, the Auxiliary and Turbine Buildings and clean soil downstream of the Turbine Building. A h ypothetical well located 2 m from the edge of the Turbine Building is shown. To address the higher contamination levels anticipated in the A uxiliary Building, the T urbine B uilding c ontamination l evel w as r educed b y a factor o f 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 2 Figure 3 Schematic Representation of Flow the geometry used to assess flow to a well outside the Turbine Building.

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TSD 14-009 Revision 2

3. Analysis Parameters All release models ar e established us ing t he un it s ource t erm a nd grounded i n c onservative 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 di ffusion r elease m odel i s us ed i n bui ldings w ith vol umetric contamination of the concrete.

3.1 Parameters Initial c onditions a ssumed t hat t he g roundwater c oncentration of e ach contaminant w as z ero everywhere. T he s ource t erm i s m odeled s uch t hat t he r esults can b e s caled t o t he a ctual inventory of t he va rious bui ldings on s ite. For t his m odeling s cenario, each b uilding w as 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. T herefore, 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 w ere s elected f or the screening m odel. W ith a ny o f the f ill ma terials it i s 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 di stribution c oefficients ( Kd) ar e i mportant p arameters in c ontrolling th e e quilibrium concentrations and transport (if modeled). A study (Sullivan, 2014) reviewed the literature and site-specific d ata t o provide c onservative va lues f or K d in a ssessing groundwater dos e. In selecting v alues from t he l iterature, e nvironmental c onditions w ith hi gh pH ( cement s orption data) as well as environmental data (soil sorption) data were considered. F or conservatism the minimum value from these conditions was selected. For nuclides with measured site-specific Kd values, the lowest measured K d 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. T o accomplish this in DUST-MS the flow velocity is set to zero.

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TSD 14-009 Revision 2 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 t he di ffusion r elease m odel t he s elected di ffusion c oefficients w ere presented in T able 4.

The ba se c ase m odel a ssumes t hat c ontamination i s uni formly di stributed ove r 0.5 inch i n t he 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 di rection of f low. D UST-MS m odel r equires a f low ar ea t o cal culate t he co rrect concentrations a bove t he f loor. T he f low a rea i s de fined a s t he a rea pe rpendicular t o t he transport direction. In these simulations, the transport direction is towards the Lake. Therefore, the f low i s t he pr oduct of t he h eight of t he w ater t able a bove t he f loor a nd t he w idth of t he building that is parallel to the Lake. T able 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 vol ume f or each b uilding. T hese w idths, height t o t he w ater t able, a nd volumes were calculated by ZionSolutions staff (Farr, 2014).

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TSD 14-009 Revision 2 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 t he ba se c ase t he f low ve locity i s s et t o z ero i n t he D UST-MS in put f ile. T o s imulate transport t o a r eceptor well soil pr operties a nd the g roundwater f low r ate a re r equired. T hese values are presented in Table 7. The K d values used were identical to those in the equilibrium model. S ite-specific s oil Kd values for C o (1161 centimeters cubed p er gram - cm3/g) and C s (527 c m3/g) a re m uch higher t han us ed i n t he a nalysis a nd t heir use w ould l ead t o l ower predicted c oncentrations. F or c onservatism, i t w as de cided t hat t he l owest K d value f rom al l sources (Sullivan, 2014) would be us ed. T he reason for us ing t he l owest Kd values is that the water l eaving t he bui lding s tructures w ould ha ve a hi gh pH due t o t he backfill m aterial. T his 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 20 of 43

TSD 14-009 Revision 2 perpendicular t o f low r emain c onstant. T herefore, f or t his s imulation onl y t he por tions of t he Turbine Building do wnstream f rom t he Auxiliary Building a re m odeled. T he l ength of t he Turbine Building p arallel t o f low i s 29.3 m . Therefore, t he t otal floor a rea of t he T urbine Building f or th is s imulation is 2 ,344 square m eters ( m2). T his i s not t he a ctual a rea of t he Turbine Building m odeled i n t he b ase case. The r eceptor w ell i s 2 m eters downstream of t he 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 bui ldings. T he actual de pth t o t he w ater t able i s d eeper i n t he Auxiliary B uilding as compared t o t he T urbine B uilding. T he geometry and flow d irection r equires t hat an y r elease from t he A uxiliary B uilding t ravel t hrough t he T urbine B uilding. T herefore, t he a ppropriate 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 pC i/m2 and the total inventory is 2554 pCi.

The i nventory of t he T urbine B uilding at th e time o f lic ense te rmination w ill b e v ery c lose to zero but is assumed to be 0.001 pCi/m2 for a total inventory of 14.7 pCi. The differences in total area l ead t o t he s lightly less than a f actor of 1 ,000 di fference i n t otal inventory i n t he t wo buildings.

3.1.4 Sensitivity Analysis Parameters To qua ntify t he i mpact of c hanges i n ke y v ariables on t he pr edicted c oncentrations a dditional calculations were performed. C haracterization d ata in dicate th at th e A uxiliary Building w ill have t he m ajority of r esidual c ontamination. F or t his r eason, a ll s ensitivity a nalyses w ill be performed f or t hat bui lding. F or s ensitivity analysis all p arameters w ere v aried b y 2 5% fro m 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 c alculating pot ential e xposures one s cenario c onsiders r emoving t he dr ill s poils f rom a hypothetical intruder well placed in the middle of the building. These drill spoils are mixed with surface s oil a nd t he r esulting dos e f rom t he c ontaminated s oil i s calculated. T he K d 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 K d values from the native sand. F or tritium (H-3) the K d value was raised from 0 to 1. Site-specific values for Europium Kd are not available. The 75th percentile value for Kd in s oils (7222 ml/g) w as us ed i n t he a nalysis ( NRC, 2000) . T able 9 l ists t he s elected K d values for the drill spoils sensitivity analysis.

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TSD 14-009 Revision 2 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 22 of 43

TSD 14-009 Revision 2 4 Results 4.1 Base Case Release Peak Groundwater Concentration Results The c onceptual m odel a ssumes t hat t he any inventory released i nstantly c omes t o e quilibrium with t he f ill m aterial t hrough t he s orption pr ocess a s c ontrolled b y t he va lue of K d. F or t he instant r elease m odel the m aximum co ncentrations o ccur at t ime = 0 before any r adioactive decay or t ransport i n t his m odel. F or t he di ffusion c ontrolled r elease, t he t ime t o t he pe ak concentration de pends o n t he di ffusion c oefficient a nd r adionuclide h alf-life. Tables 9 - 14 provide t he m aximum c oncentration i n e ach bui lding. T he tables a lso provide the a mount of radioactivity ( pCi) i n s olution, t he a mount s orbed t o t he s olid m aterial ( pCi) a nd t he concentration on the fill material (pCi/g) with a density of 1.5 g/cm3.

4.1.1 Auxiliary Building The ba se c ase f or t he A uxiliary Building a ssumes a di ffusion c ontrolled r elease. 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 6467 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 a nd a hi gh di ffusion r ate r eleases a lmost a ll t he i nventory within t he f irst year t o solution. Sr-90 w ith t he l ow K d value of 2.3 s hows s lightly more than 4% (140.1 pCi) of t he total inventory (6503 pCi) is i n s olution. For all o ther n uclides th e ma ximum a ctivity in th e water is less than 0.2% of the entire inventory. For Ni-63 the peak activity sorbed to the solid (5051 pCi) is s lightly le ss th an 80% of t he t otal a ctivity ( 6503 pCi). T his r eflects th e time -

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TSD 14-009 Revision 2 dependent r elease f rom the co ncrete an d t he ef fects o f r adioactive d ecay. The time to p eak represents the balance between the release rate, sorption, and radioactive decay. The value in the table is a pproximate as the c oncentration s hows a br oad pe ak ov er t ime. T he r adionuclides 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. I n determining the potential dose, the results of this a nalysis w ill be s caled b y t he m easured i nventory i n e ach bui lding. T he C ontainment Buildings will have all of t he concrete i nside t he l iner 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 C rib H ouse/Forebay i s expected t o contain little or no c ontamination ba sed on characterization data and the contamination that may be present will be at the surface. F or this reason, the instant release model was used. Table 12 provides the results of the analysis.

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TSD 14-009 Revision 2 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 F uel P ool a nd T ransfer C anals has n ot b een f ully c haracterized a t th is time . It is believed that there will be volumetric contamination in the concrete below the pool liners. For this r eason di ffusion c ontrolled r elease i s m odeled assuming uni form c ontamination i n t he 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 i mpact of di ffusion controlled r elease on pe ak c oncentrations i s s lightly m ore pr onounced than in the Auxiliary Building with a peak solution concentration for Sr-90 slightly in excess of 2 percent of t he t otal i nventory. The H -3 c oncentration pr edicted f or t he Fuel Building ( 0.015 19 Page 25 of 43

TSD 14-009 Revision 2 pCi/L) is the highest predicted concentration for any of the buildings. T his is due to the small amount o f w ater a vailable f or mix ing and t he hi gh di ffusion r elease r ate ( over 9 9% of t he 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 c ontamination based on c haracterization data and any contamination t hat may b e pr esent w ould b e on t he s urface. For t his r eason t he i nstant 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 a ll of t he bui ldings with t he e xception of H -3. T he c ause for t his i s t he ve ry l ow m ixing 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 di rectly proportional to s urface a rea. T herefore, a hi gh s urface area t o vol ume r atio w ill 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 2 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 s ensitivity an alysis was p erformed o n t he k ey parameters i n t he b ase cas e m odel for t he Auxiliary Building. T he 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 c ases. A ppendix A c ontains t he de tailed r esults of t hese s imulations a nd i ncludes T ables 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). A dditionally, 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 a ll nuc lides except H -3, most o f th e a ctivity is s orbed o nto th e b ackfill ma terial.

Strontium w ith th e lo west K d still ha d m ore t han 90% of t he a ctivity sorbed on t he backfill.

• Kd: An increase in Kd caused a decrease in solution concentration and a slight increase in sorbed concentration. S olution concentration is approximately inversely proportional to Kd. T he 25% change in K d had a minimal impact on t he amount sorbed or the backfill concentration ( pCi/g). S trontium showed t he l argest pe rcentage c hange i n s orbed concentration of all the nuclides but it was less than 2.5%.

• Porosity: C hanging por osity h ad a m inor i mpact on t he a mount s orbed a nd s olution concentration. T he amount of radioactivity in solution was proportional to the porosity.

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TSD 14-009 Revision 2 This r eflects t he av ailability o f w ater w ith h igher p orosity h aving m ore w ater av ailable for mixing and a higher total amount of activity in the water.

• Density: T he s olution c oncentration, s orbed c oncentration and amount i n s olution are inversely p roportional t o de nsity. Increasing density c auses a de crease i n s olution concentration. T he change in density has a minor impact (< 2%) on t he 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 t hick c ontaminated z one. In all c ases, t he t otal i nventory for e ach nuc lide r emained constant a t 6503 pC i. Changes i n t he de pth of c ontamination c an l ead to c hanges i n t he t otal amount of mass released, the peak concentration, and the time to reach the peak concentration.

Table 1 6 examines t he i mpact of contaminated zone t hickness on t he p ercentage o f t he t otal inventory released into solution over time and compares the change in total mass released to the base c ase 1 /2 inch t hick c ontaminated z one. H-3 has t he hi ghest di ffusion c oefficient a nd releases ove r 98% of t he i nventory i n a ll t hree c ases a nd t herefore, t he c ontaminated z one thickness only has a minor impact on the total mass released. The nuclides with a short half-life or a l ow di ffusion c oefficient in th is s imulation ( Co-60, Sr-90, Cs-134, E u-152, a nd E u-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. T he 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 28 of 43

TSD 14-009 Revision 2 Table 17 provides the peak water concentration as a function of contaminated zone thickness and the p ercentage change from t he b ase c ase ( 0.5 inch t hick c ontaminated z one). The p eak concentrations followed the same trends as the p ercentage 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 i nches. T he ot her nuc lides s howed a n a lmost l inear r esponse w ith c ontamination 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 t o r each t he p eak concentration i s a b alance b etween t he d iffusion r elease r ate an d t he radioactive d ecay r ate. H-3 is v ery s ensitive to c ontaminated z one th ickness in th e 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. H owever, 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 2 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 di scussed i n s ection 3.1.3 one e xposure s cenario i ncludes us ing t he drill s poils a nd m ixes them with the native soil. T o examine the change in drill spoils radionuclide concentration the Kd values i n T able 9 w ere u sed. T able 19 pr ovides t he r esults f or t he new K d values in t he 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 2 0 co mpares t he sensitivity case t o t he b ase cas e f or t he p eak concentration an d p eak 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 30 of 43

TSD 14-009 Revision 2 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 t ime e volution of c oncentration a t a r eceptor w ell l ocated t wo m eters out side t he T urbine Building w as s imulated u sing th e b ackfill material K d values i n T able 4, t he s oil K d and groundwater p arameters i n T able 7, a nd t he g eometry i n F igure 3. The in itial 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). T his 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, di ffusion controlled release i s assumed for t he Auxiliary Building and Instant R elease i s assumed for the Turbine Building.

Table 18 provides the peak concentration in the Auxiliary Building, Turbine Building, Edge of the T urbine B uilding, and t he R eceptor W ell. T o qua ntitatively de fine t he r eduction i n 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 t he R eceptor W ell i s also provided. Recalling th at the initial in ventory 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 b y around a factor of ten in traveling two meters from t he e dge of t he T urbine B uilding t o t he Receptor W ell. H-3, th e mo st mo bile n uclide 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 31 of 43

TSD 14-009 Revision 2 of 0.78 after 23 years, slightly less than that for the more mobile H-3. A ll 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 uppe r boun d on pe ak c ontaminant c oncentrations. The f ollowing q ualitative arguments support this assertion.

• The R easonably Foreseeable Scenario, d efined i n N UREG 1757 a s a l and us e s cenario 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.

26 Page 32 of 43

TSD 14-009 Revision 2 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:

= /[ ( + )]

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 co mparison w as m ade b etween t he D UST-MS out put a nd t he a nalytical s olution i n t he equation a bove for t he Turbine B uilding as an example o f an i nstant r elease b asement. Th e 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 33 of 43

TSD 14-009 Revision 2 6 Conclusions A model for predicting peak groundwater concentrations at the ZSRP Site after decommissioning has b een developed. T he m odel us es t he D UST-MS s imulation mo del w hich c alculates th e release a nd t ransport of r adioactive c ontamination i n a groundwater s ystem. T he a nalysis i s based on a unit source term of 1 pCi/m2 on the entire wall and floor surface area of each of the seven bui ldings t hat will ha ve a residual be low gr ound, ba ckfilled s tructure. Conservative assumptions ba sed on existing da ta w ere us ed i n t he s creening m odel for selecting p arameters that i mpact groundwater co ncentration (Kd, porosity, bul k density, no f low). For example, t he Kd value s elected f or t he f ill m aterial w as t he l owest m easured v alue using s ite-specific groundwater for a ny soil o r f ill ma terial. The r esults of t he m odel c an be c ombined w ith 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 t he A uxiliary Building ba se case. The results o f t he an alysis showed t hat t he pe ak 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 t he release model through comparison of the diffusion change in total mass released, peak concentration, and time to reach the peak concentration for the ba se case, on e i nch contaminated z one, t o r esults f rom s imulations w ith one-half and t wo 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 m ass w as r eleased i n al l t hree s imulations. The ot her nuclides s howed close t o an i nverse linear d ependence o n co ntaminated z one l ength with t he m ass r elease a nd p eak co ncentration decreasing by close to a f actor of two with an increase in length of a f actor of two. T he time to reach the peak concentration was independent of the length of the contaminated zones for short-lived nuc lides ( other t han H -3) i ndicating t hat a b alance b etween r elease r ate an d radioactive decay was achieved. F or H-3 the high release rate caused the peak concentration to be reached in 0.1 years f or t he s hortest c ontaminated l ength ( 1/2 i nch) and 1.4 years f or t he t wo i nch contaminated length simulation.

Removing t he a ssumption of a w ell placed in t he mid dle o f th e f ill ma terial a nd p lacing th e Receptor Well two m eters out side t he T urbine Building, w hich i s t he closest s oil ( e.g. non -

building) location to th e A uxiliary Building w here th e h ighest r esidual c ontamination w ill 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 o f fi fty reduction f or N i-63. H-3 s howed a 20% r eduction i n pe ak dos e 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 34 of 43

TSD 14-009 Revision 2 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 & A ssociates, 201 4, Evaluation of Hydrological Parameters in S upport 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 . S ullivan ( 2014). Sorption ( Kd) measurements on C inder B lock and G rout i n Support of Dose Assessments for Zion Nuclear Station Decommissioning, Brookhaven National Laboratory Report to ZionSolutions, April 2014.

Muurinnen, A , J. R antanen, R . O vaskainen and O .J. H einonen, Diffusion M easurements i n 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 D iffusion C oefficients f or R adionuclides i n C ement, personal communication.

Sullivan, T.M., "DUST - Disposal Unit Source T erm: 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 U nit Source T erm - Multiple S pecies - 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 f or D ecommissioning the Zion N uclear P ower Plant, R evision 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 35 of 43

TSD 14-009 Revision 2 Nuclear Chemistry, Vol. 252, No. 1 (2002) 133-138.

U.S. Nuclear Regulatory Commission, (NRC, 2000). Development of Probabilistic RESRAD 6.0 and R ESRADBUILD 3 .0 C omputer C odes, NUREG/CR-6697, U.S. N uclear R egulatory Commission, December 2000.

Yim, S .P, T .M. S ullivan, a nd L. M ilian, Sorption ( Kd) m easurements i n S upport of D ose Assessments f or Zion Nuclear S tation D ecommissioning, B rookhaven N ational Laboratory Report to ZionSolutions, December 12, 2012.

30 Page 36 of 43

TSD 14-009 Revision 2 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 nuc lides ( H-3) a nd s orbing nuc lides. T he non -sorbing nu clide s howed a pproximately 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 Peak Peak Sorbed Coefficient Kd to Concentration Radioactivity Radioactivity Concentration Peak in Solution Sorbed 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 37 of 43

TSD 14-009 Revision 2 A.2: High Kd Kd values ar e i n t he t able b elow. T hey w ere i ncreased b y 2 5% f rom t he b ase cas e v alue.

Increasing the K d 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 Coefficient Kd Concentration Radioactivity in Radioactivity 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 38 of 43

TSD 14-009 Revision 2 A.3: Low Kd Kd values a re s hown i n the t able below and w ere r educed b y 2 5% f rom t he b ase case v alues.

Reducing K d 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 39 of 43

TSD 14-009 Revision 2 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. F or 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 40 of 43

TSD 14-009 Revision 2 A.5: Low Porosity The por osity w as de creased t o 0.19 from t he b ase c ase v alue of 0.25. F or s orbing nuc lides decreasing t he por osity did not i mpact t he s olution c oncentration but i t di d r educe t he t otal 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 41 of 43

TSD 14-009 Revision 2 A.6: High Bacfkill Density The ba ckfill de nsity w as i ncreased t o 1.8 g/cm3 from t he b ase case v alue o f 1 .5 g/cm3.

Increasing t he de nsity caused bot h t he s olution c oncentration a nd s orbed c oncentration t o decrease for sorbing nuclides. T his 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 42 of 43

TSD 14-009 Revision 2 A.7: Low Density The d ensity w as d ecreased t o 1 .1 g/cm3 from t he b ase case v alue o f 1 .5 g /cm3. R educing t he density caused an increase in both the solution concentration and the sorbed concentration. The increase was i nversely p roportional t o t he density. T he change i n density did not i mpact 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 43 of 43