ML16211A379

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TSD-14-031 Rev. 1, Brookhaven National Laboratory: Evaluation of Maximum Radionuclide Groundwater Concentrations for Basement Fill Model
ML16211A379
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Site: Zion  File:ZionSolutions icon.png
Issue date: 07/20/2016
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Office of Nuclear Material Safety and Safeguards
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ZS-2016-0084 TSD-14-031, Rev. 1
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ZionSolutions, LLC.

Technical Support Document TSD 14-009

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ZIONSOLUTIONSuc 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: __ ---=:.~ __

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BNL-107249-2014-IR-R1 Basement Fill Model Evaluation of Maximum Radionuclide Concentrations for Initial Suite of Radionuclides Zion Station Restoration Project Terry Sullivan Brookhaven National Laboratory Revision 1 May 20, 2016 TSD 14-031 Revision 1 Page 2 of 25

Revision Log Section Page Rev.

Date Reason(s) for Revision Title page 1

5/20/16 Revised Revision number Title page 1

5/20/16 Revised effective date 3.2.1 12, Table 7

1 5/20/16 Changed the total pCi released for H-3 in Table 7 from 6503 pCi to 6467 pCi. The value 6503 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.

3.2.3 14, Table 9

1 5/20/16 Fixed typo error for Ni-63 peak concentration value. The correct value, 2.44E-6 pCi/L now matches the value for Ni-59.

TSD 14-031 Revision 1 Page 3 of 25

Table of Contents 1 Introduction.................................................................................................................................. 1 2 Conceptual Models of Release..................................................................................................... 2 2.1 Site Overview........................................................................................................................ 2 2.2 Modeling Overview.............................................................................................................. 4 2.3 Release Models..................................................................................................................... 7 2.3.1 Instant Release............................................................................................................... 7 2.3.2 Release Rate: Diffusion Controlled Release from the concrete.................................... 7 3 Analysis........................................................................................................................................ 8 3.1) Parameters........................................................................................................................... 9 3.1.1 Diffusion Controlled Release Model........................................................................... 10 3.1.2 Model Geometry.......................................................................................................... 10 3.2 Peak Groundwater Concentration Results......................................................................... 11 3.2.1 Auxiliary Building....................................................................................................... 11 3.2.2 Containment Building.................................................................................................. 12 3.2.3 Crib House/Forebay Building...................................................................................... 14 3.2.4 Turbine Building.......................................................................................................... 15 3.2.5 Spent Fuel Building..................................................................................................... 16 3.2.6 Waste Water Treatment Facility.................................................................................. 17 4.0 Conclusions............................................................................................................................. 18 5.0 References............................................................................................................................... 19 TSD 14-031 Revision 1 Page 4 of 25

Figures Figure 1 Zion Site layout for buildings that will have a residual underground structure............... 3 Tables Table 1 Mixing volume and release rate assumption..................................................................... 5 Table 2 Geometric Parameters and Unit Inventory for Residual Structures (Farr, 2014)............. 6 Table 3. Potential Radionuclides of Concern at the Zion Power Plant.......................................... 7 Table 4 Typical diffusion coefficients in cement for radionuclides of concern............................ 8 Table 5 Selected distribution coefficients (Sullivan, 2014) and diffusion coefficients............... 10 Table 6 Model Geometry for all simulations............................................................................... 11 Table 7 Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2 in the Auxiliary Building......................................................................................................................................... 12 Table 8 Peak Groundwater and Soil Concentrations per unit source of 1 pCi/m2 in the Containment Building................................................................................................................... 13 Table 9 Peak Groundwater and Soil Concentrations per unit source of 1 pCi/m2 in the Crib House/Forebay Building............................................................................................................... 14 Table 10 Peak Groundwater and Soil Concentrations per unit source of 1 pCi/m2 in the Turbine Building......................................................................................................................................... 15 Table 11 Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2 in the Spent Fuel Building................................................................................................................................ 16 Table 12 Peak Groundwater and Soil Concentrations per unit source of 1 pCi/m2 in the Crib House/Forebay Building............................................................................................................... 17 TSD 14-031 Revision 1 Page 5 of 25

1 Basement Fill Model Evaluation of Maximum Radionuclide Concentrations for Initial Suite of Radionuclides 1 Introduction ZionSolutions is in the process of decommissioning the Zion Nuclear Power Plant. The current decommissioning plan involves removing all above grade structures to a depth of 3 feet below grade. T he remaining underground structures will be backfilled with clean material. The final selection of fill material has not been made. The remaining backfilled structures will be the two reactor Containment Buildings, a S pent Fuel B uilding, an Auxiliary Building, a Turbine Building, a Crib House/Forebay Building and a Waste Water Treatment Facility (WWTF).

Remaining structures will contain low amounts of residual radioactive material. The bulk of the source term will be contained in the concrete floors which are twenty to thirty feet below grade in most buildings. Current interior demolition plans are to remove all concrete in the Unit 1 and Unit 2 R eactor Buildings inside the steel liner. Based upon concrete characterization data, the highest e nd s tate s ource t erm i s a nticipated t o be contained in th e A uxiliary Building floor located approximately 50 feet below grade. Thus the end state source term will be well below grade and below the water table eliminating conventional pathways such as direct radiation and inhalation r endering groundwater r elated p athways t he m ost s ignificant pot ential s ources of future exposure. Note that the Spent Fuel Pool concrete under the liner is also expected to be significantly contaminated but characterization has not yet been performed due to inaccessibility.

The floor of the spent fuel pool is 15 feet below grade.

In order to terminate the Part 50 l icense, the Zion Solutions Restoration Project (ZSRP) must demonstrate that the dose from remaining residual radioactivity does not cause a h ypothetical individual to receive a dose in excess of 25 mrem/y-1 as specified in 10 CFR 20 Subpart E. To demonstrate co mpliance, t he modeling of t he fate and t ransport of r adioactive m aterial t o a receptor is r equired. For t he ba ckfilled ba sements, t his mo deling th e r elease o f r adioactivity from the concrete and mixing with the water contained in the fill material. This report determines the ma ximum groundwater concentrations in t he b asement f ill f or an in itial s uite o f 2 6 radionuclides designated by the ZSRP as having a potential of being present on the Zion Nuclear Power Station (ZNPS). The results of this analysis are used by ZSRP to determine the relative dose contribution from all 26 radionuclides to identify the insignificant dose contributors and select the radionuclides of concern (ROCs) to be included in more detailed calculations.

This report uses the same methods described in detail in the DUST Report (Sullivan, 2014) with the addition of parameters required for the additional radionuclides evaluated in this report. The applicable p arts o f the c onceptual m odel a nd i nput pa rameter de scriptions f rom the DUST Report are repeated for completeness in order to allow this report to be a stand alone document.

Calculation of the release of radioactive material from the remaining building basements requires site-specific i nformation on t he h ydrogeologic t ransport pr operties ( effective por osity, bul k density, hydraulic) and chemical transport properties (sorption). Conestoga-Rovers & Associates (CRA) ha s c ollected a substantial a mount of s ite-specific h ydrogeologic da ta ( CRA, 201 4).

TSD 14-031 Revision 1 Page 6 of 25

2 However, this screening calculation estimates only the water concentration in the basement fill.

No transport away from the basements is assumed, which would result in lower concentrations.

Brookhaven National Laboratory has determined site-specific sorption data for five nuclides and four soil types, two concrete construction demolition debris, two cinder block materials, and one grout material that are under consideration for the fill (Yim, 2012, Milian, 2014). In addition, sand from the local region could be used as part or all of the backfill. The composition of the fill material has not been finalized.

The objectives of this report are:

a) To pr esent a s implified c onceptual m odel f or release f rom t he bui ldings w ith r esidual subsurface s tructures t hat c an b e us ed t o p rovide a n uppe r bound o n radionuclide concentrations in the fill material and the water in the interstitial spaces of the fill.

b) Provide maximum water concentrations and the corresponding amount of mass sorbed to the s olid f ill ma terial th at c ould o ccur in e ach building f or u se b y ZSRP in s electing ROCs for detailed dose assessment calculations.

2 Conceptual Models of Release 2.1 Site Overview Figure 1 provides t he s ite l ayout a t Z NPS l ocated on t he s hores of Lake M ichigan. M ajor features i nclude t wo r eactor C ontainment B uildings ( Unit-1 a nd U nit-2), a F uel H andling Building, A uxiliary Building, T urbine B uilding, C rib H ouse, a nd W aste W ater T reatment Facility (WWTF).

The pr oposed de commissioning a pproach i nvolves r emoval of r egions w ith hi gh-levels o f 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 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. T he 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 w ill 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 TSD 14-031 Revision 1 Page 7 of 25

3 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.

Figure 1 Zion Site layout for buildings that will have a residual underground structure.

TSD 14-031 Revision 1 Page 8 of 25 F~ I

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4 2.2 Modeling Overview The D isposal 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 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). 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.

The c onceptual mo del f or r elease is imp ortant f or d etermining th e a mount o f ma terial in th e water and fill. In many 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 th at th e e ntire in ventory is a vailable imme diately a fter lic ense te rmination. In s ome buildings t he contamination i s e xpected t o h ave di ffused i nto t he concrete r esulting i n 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.

A s econd i mportant p arameter i s t he v olume o f w ater av ailable t o m ix w ith r eleased radionuclides. Table 1 s ummarizes th e to tal f ill v olume a vailable f or mixing a nd th e r elease 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. T he 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 w ill be difficult to achieve this packing density. For example, the native sand has total porosity greater than 30%.

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 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. D iffusion 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 modeled as an Instant Release.

TSD 14-031 Revision 1 Page 9 of 25

5 Table 1 Mixing volume and release rate assumption Building Volume* (m3)

Release Rate Assumption Unit 1 Containment 6.537E+03 Instant Release (loose surface contamination)

Unit 2 Containment 6.537E+03 Instant Release (loose surface contamination)

Auxiliary 2.84E+04 Diffusion Controlled Release (concrete contamination at depth in concrete)

Turbine 2.61E+04 Instant Release (the limited contamination present is at the concrete surface with very limited contamination at depth.)

Crib House and Forebay 3.05E+04 Instant Release (limited or no surface contamination)

Waste Water Treatment Facility 1.44E+02 Instant Release (limited or no surface contamination)

Spent Fuel Pool and Transfer Canals 2.08E+02 Diffusion Controlled Release (Concrete contamination expected at depth under the liner)

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 t he concrete surfaces and the Instant Release model is used.

Diffusion C ontrolled R elease w as u sed t o es timate t he s ource t erm r elease r ate f or t he Fuel Handling Building Basement and Fuel Transfer Canals due to expected contamination at depth in concrete after the liners are removed.

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 width was adjusted to match the total volume in Table 1. 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.

The inventory for each building was based on a uniform contamination level for each nuclide of 1 pC i/m2 on t he w all a nd f loor s urfaces. T his c ontamination l evel w as us ed f or m odeling 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 c ompleted. Table 2 c ontains t he building s urface a reas f or t he c alculations of inventory. From Table 2 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.

TSD 14-031 Revision 1 Page 10 of 25

6 Table 2 Geometric Parameters and Unit Inventory for Residual Structures (Farr, 2014)

Structure Basement Floor Elevation (feet)

Distance to Water Table meters Structure Total Surface Area (m2)

Inventory (Ci)

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 Gen Oil Storage 560 5.79 14679 1.468E-08 Spent Fuel Pool and Transfer Canals 576 0.91 780 7.80E-10 Waste Water Treatment Facility 577 0.61 1124 1.124E-09 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) 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 Kd for dose assessment was p erformed (Sullivan, 2014). The K d values s elected f rom t he l iterature w ere chosen t o give a co nservative es timate o f w ater co ncentration ( highest v alue) f or d ose assessment. W hen site-specific values are available, the lowest K d value measured in any fill material or soil was selected.

The compliance assessment requires prediction of the release and transport of contaminants to the h ypothetical i ndividual. C haracterization s tudies a nd a ssessments by Z ionSolutions have identified the following potential radionuclides (Table 3). All nuclides in Table 3 were used in the simulation of maximum groundwater concentration.

TSD 14-031 Revision 1 Page 11 of 25

7 Table 3. Potential Radionuclides of Concern at the Zion Power Plant Radionuclides H-3 C-14 Fe-55 Ni-59 Co-60 Ni-63 Sr-90 Nb-94 Tc-99 Ag-108m Sb-125 Cs-134 Cs-137 Pm-147 Eu-152 Eu-154 Eu-155 Np-237 Pu-238 Pu-239/240 Pu-241 Am-241 Am-243 Cm-243/244 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 ha ve be en conducted f or t he di ffusion i n c oncrete o f t he r adionuclides und er consideration at Zion (H-3, Co-60, Ni-63, Sr-90, Cs-134, Cs-137, Eu-152, and Eu-154). T he TSD 14-031 Revision 1 Page 12 of 25

8 diffusion coefficient from concrete will depend on the water to cement ratio used in forming the concrete and the aggregate. A t ypical range from the literature is presented in Table 4. T he 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 Range (cm2/s)

Selected Diffusion Coefficient (cm2/s)

Reference 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 Serne (Serne, 2001) provided a table of best estimates for diffusion coefficients for forty-four elements relevant for nuclear waste disposal including most of the elements listed in Table 3. In general, t he r ecommended d iffusion co efficient f or m ost el ements w as l ess t han t he va lue recommended for Cs. C s is known to be relatively mobile in cement systems and in fact, with the exception of H-3, has the highest recommended value for diffusion coefficient in Table 4.

The el ements w ith d iffusion co efficients g reater t han C s i ncluded e lements t hat t end t o f orm anionic species (such as Tc and I). Based on this information, all elements in Table 3 without a specific diffusion coefficient in Table 4 are set to the value used for Cs with the exception of Tc, which i s assigned t he s ame d iffusion co efficient as H -3. T his s hould p rovide a conservative upper bound on release of these species from the Auxiliary and Fuel Buildings. For the other buildings the release is instantaneous and the diffusion coefficient does not impact release.

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 i s uniformly contaminated over a 0.5 i nch thickness and that all of the material is released at the 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. 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 f low boundary and this is equivalent to having a s lab 0.5 i nch t hick but pr eventing di ffusion f urther i nto t he c ement. T his i s 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.

TSD 14-031 Revision 1 Page 13 of 25

9 3 Analysis All r elease m odels a re 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 c oncentration. 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, e ach bui lding 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. Therefore, the bulk density and porosity are unknown. A bulk density of 1.5 grams per cubic centimeter (g/cm3) and an effective porosity o f 0.25 w ere s elected f or th e s creening mo del. 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) a re important 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 p rovide co nservative v alues 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 Kd in any backfill or soil was selected. Selected values are in Table

5. T able 5 a lso provides the diffusion coefficient used to simulate release from volumetrically contaminated concrete in the Auxiliary Building and the Fuel Building.

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.

TSD 14-031 Revision 1 Page 14 of 25

10 Table 5 Selected distribution coefficients (Sullivan, 2014) and diffusion coefficients Radionuclide Half Life (years)

Basement Fill Kd to Be Used ml/g Recommended Diffusion Coefficient (cm2/s)

H-3 12.4 0

5.5E-7 C-14 5730 1.2 3.0E-9 Fe-55 2.7 2857 3.0E-9 Ni-59 75000 62 1.1E-9 Co-60 5.27 223 4.1E-11 Ni-63 96 62 1.1E-9 Sr-90 29.1 2.3 5.2E-10 Nb-94 20300 45 3.0E-9 Tc-99 213000 0

5.5E-7 Ag-108m 127 27 3.0E-9 Sb-125 2.77 17 3.0E-9 Cs-134 2.06 45 3.0E-9 Cs-137 30 45 3.0E-9 Pm-147 2.62 95 3.0E-9 Eu-152 13.3 96 5.0E-11 Eu-154 8.8 96 5.0E-11 Eu-155 4.96 96 5.0E-11 Np-237 2140000 1

3.0E-9 Pu-238 87.7 174 3.0E-9 Pu-239 24100 174 3.0E-9 Pu-240 65400 174 3.0E-9 Pu-241 14.4 174 3.0E-9 Am-241 432 177 3.0E-9 Am-243 7380 177 3.0E-9 Cm-243 28.5 891 3.0E-9 Cm-244 18.1 891 3.0E-9 3.1.1 Diffusion Controlled Release Model For t he di ffusion release model t he s elected d iffusion co efficients w ere presented i n T able 5.

The base case model assumes that contamination is uniformly distributed over 0.5 i nch 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 di rection of f low. D UST-MS m odel r equires a f low ar ea to 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 TSD 14-031 Revision 1 Page 15 of 25

11 transport direction. In these simulations, the transport direction is towards the Lake. Therefore, the flow i s t he product of t he h eight of t he w ater t able above t he floor and t he width of t he 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 vol ume f or each b uilding. T hese w idths, height t o t he w ater t able, a nd vol umes w ere calculated by ZionSolutions staff (Farr, 2014).

Table 6 Model Geometry for all simulations.

Structure Width or Radius m

Height to Water Table m

Flow Area (m2)

Contaminated Zone Length (m)

Void Space to WT m3 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 Facility 14.63 0.61 8.919 16.09 144 Spent Fuel Pool and Transfer Canals 10.06 0.91 18.64 11.17 208 3.2 Peak Groundwater Concentration Results 3.2.1 Auxiliary Building The conceptual model assumes that the inventory is released through diffusion through a one-half i nch uni formly c ontaminated z one on t he floor. T he nu clides r eleased i nto t he w ater instantly reach equilibrium with the fill material through the sorption process as controlled by the value of Kd. T he results of this model are presented in Table 7. In addition to the maximum groundwater concentration, the table provides the amount of radioactivity (pCi) in solution, the amount sorbed to the solid material (pCi) and the peak concentration sorbed on the soil (pCi/g).

The t otal i nventory w as 6.503E-09 Ci (6503 pCi) for e ach nu clide m odeled. T he vol ume between the floor and the water table is 28445 m3, Table 1.

Examining Table 7 the impact of sorption is clear. For example, consider Sr-90 with a Kd of 2.3 ml/g, the solution concentration is less than 3% of the value for Kd = 0. Table 7 shows the impact of diffusion and decay on release as the sum of the radioactivity in the water and soil is often substantially less than the total inventory of 6503 pCi.

TSD 14-031 Revision 1 Page 16 of 25

12 Table 7 Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2 in the Auxiliary Building Half-life Kd Diffusion Coefficient Peak Concentration Peak Radioactivity in Solution Peak Radioactivity Sorbed Peak Sorbed Concentration Nuclide (years)

(ml/g)

(cm2/s) pCi/L pCi pCi pCi/g H-3 12.4 0

5.5E-07 9.10E-04 6467 0

0 C-14 5730 1.2 3.0E-09 1.11E-04 793 5710 1.33E-07 Fe-55 2.7 2857 3.0E-09 1.24E-08 0.09 1519 3.54E-08 Ni-59 75000 62 1.1E-09 2.45E-06 17.4 6512 1.52E-07 Co-60 5.27 223 4.1E-11 2.6E-08 0.2 249 5.80E-09 Ni-63 96 62 1.1E-09 1.90E-06 13.6 5051 1.18E-07 Sr-90 29.1 2.3 5.2E-10 1.96E-05 140.1 1933 4.51E-08 Nb-94 20300 45 3.0E-09 3.38E-06 24 6521 1.52E-07 Tc-99 213000 0

5.5E-07 9.15E-04 6503 0

0 Ag-108m 127 27 3.0E-09 5.23E-06 37 6054 1.41E-07 Sb-125 2.77 17 3.0E-09 2.08E-06 15 1516 3.54E-08 Cs-134 2.06 45 3.0E-09 6.89E-07 5

1329 3.10E-08 Cs-137 30 45 3.0E-09 2.47E-06 17.7 4766 1.11E-07 Pm-147 2.62 95 3.0E-09 3.68E-07 3

1499 3.50E-08 Eu-152 13.3 95 5.0E-11 1.07E-07 0.8 440 1.02E-08 Eu-154 8.8 95 5.0E-11 8.38E-08 0.6 341 7.96E-09 Eu-155 4.96 95 5.0E-11 6.39E-08 0

260 6.07E-09 Np-237 2140000 1

3.0E-09 1.31E-04 936 5616 1.31E-07 Pu-238 87.7 174 3.0E-09 7.84E-07 6

5848 1.36E-07 Pu-239 24100 174 3.0E-09 8.75E-07 6

6527 1.52E-07 Pu-240 65400 174 3.0E-09 8.74E-07 6

6519 1.52E-07 Pu-241 14.4 174 3.0E-09 4.78E-07 3

3566 8.32E-08 Am-241 432 177 3.0E-09 8.42E-07 6

6389 1.49E-07 Am-243 7380 177 3.0E-09 8.60E-07 6

6526 1.52E-07 Cm-243 28.5 889 3.0E-09 1.24E-07 1

4736 1.10E-07 Cm-244 18.11 889 3.0E-09 1.04E-07 1

3973 9.27E-08 3.2.2 Containment Building The conceptual model assumes that the inventory is released instantly to the water and instantly reaches equilibrium with the fill material through the sorption process as controlled by the value of K d. T he results o f t his m odel are pr esented i n T able 8. In a ddition t o t he m aximum groundwater concentration, the table provides the amount of radioactivity (pCi) in solution, the amount sorbed to the solid material (pCi) and the peak concentration sorbed on the soil (pCi/g).

The t otal i nventory f or e ach m odeled nuc lide i n t he C ontainment B uilding a ssuming a TSD 14-031 Revision 1 Page 17 of 25

13 contamination level of 1 pCi/m2 on the floors and walls is 2759 pC i. The volume between the floor and the water table is 6537 m3, Table 1.

Table 8 Peak Groundwater and Soil Concentrations per unit source of 1 pCi/m2 in the Containment Building Half-life (years)

Kd Peak Concentration (pCi/L)

Activity in Solution Activity Sorbed Peak Concentration (pCi/g)

H-3 12.4 0

1.69E-03 2759 0

0 C-14 5730 1.2 2.06E-04 336.6 2422.4 2.47E-07 Fe-55 2.7 2857 9.82E-08 0.2 2758.8 2.81E-07 Ni-59 75000 62 4.53E-06 7.4 2751.6 2.81E-07 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 Nb-94 20300 45 6.23E-06 10.2 2748.8 2.8E-07 Tc-99 213000 0

1.69E-03 2759 0

0 Ag-108m 127 27 1.04E-05 16.9 2742.1 2.8E-07 Sb-125 2.77 17 1.64E-05 26.7 2732.3 2.79E-07 Cs-134 2.06 45 6.23E-06 10.2 2748.8 2.8E-07 Cs-137 30 45 6.23E-06 10.2 2748.8 2.8E-07 Pm-147 2.62 95 2.95E-06 4.8 2754.2 2.81E-07 Eu-152 13.3 95 2.95E-06 4.8 2754.2 2.81E-07 Eu-154 8.8 95 2.95E-06 4.8 2754.2 2.81E-07 Eu-155 4.96 95 2.95E-06 4.8 2754.2 2.81E-07 Np-237 2140000 1

2.41E-04 394.0 2365.0 2.41E-07 Pu-238 87.7 174 1.62E-06 2.6 2756.4 2.81E-07 Pu-239 24100 174 1.62E-06 2.6 2756.4 2.81E-07 Pu-240 65400 174 1.62E-06 2.6 2756.4 2.81E-07 Pu-241 14.4 174 1.62E-06 2.6 2756.4 2.81E-07 Am-241 432 177 1.59E-06 2.6 2756.4 2.81E-07 Am-243 7380 177 1.59E-06 2.6 2756.4 2.81E-07 Cm-243 28.5 891 3.15E-07 0.5 2758.5 2.81E-07 Cm-244 18.11 891 3.15E-07 0.5 2758.5 2.81E-07 The peak ground water concentrations in the Containment Building are predicted to be greater than in the Auxiliary Building for a unit contamination level of 1 pCi/m2. This reflects the lower amount of water available for mixing and the higher release rate in the Containment Building.

With t he e xception of t he f ive nu clides ( H-3, C -14, S r-90, T c-99, and N p-237) t hat h ad a distribution coefficient value of less than 3, m ore than 99% of the activity released was sorbed on the solid backfill. This is the cause for the soil concentrations being similar for most nuclides.

TSD 14-031 Revision 1 Page 18 of 25

14 3.2.3 Crib House/Forebay Building The conceptual model assumes that the inventory is released instantly similar to the Containment Building. The results of t his m odel are presented i n Table 9. In addition t o t he m aximum groundwater concentration, the table provides the amount of radioactivity (pCi) in solution, the amount sorbed to the solid material (pCi) and the peak concentration sorbed on the soil (pCi/g).

The t otal i nventory f or e ach m odeled nuc lide i n t he C ontainment B uilding a ssuming a contamination level of 1 pC i/m2 on the floors and walls is 6940 pCi. T he volume between the floor and the water table is 30524 m3, Table 1.

Table 9 Peak Groundwater and Soil Concentrations per unit source of 1 pCi/m2 in the Crib House/Forebay Building Half-life (years)

Kd Peak Concentration (pCi/L)

Activity in Solution Activity Sorbed Peak Concentration (pCi/g)

H-3 12.4 0

9.08E-04 6936 0

0 C-14 5730 1.2 1.11E-04 845.8 6094.2 1.33E-07 Fe-55 2.7 2857 5.29E-08 0.4 6939.6 1.52E-07 Ni-59 75000 62 2.44E-06 18.6 6921.4 1.51E-07 Co-60 5.27 223 6.78E-07 5.2 6934.8 1.51E-07 Ni-63 96 62 2.44E-06 14.6 6925.4 1.51E-07 Sr-90 29.1 2.3 6.14E-05 468.5 6471.5 1.41E-07 Nb-94 20300 45 3.35E-06 25.6 6914.4 1.51E-07 Tc-99 213000 0

9.09E-04 6936 0

0 Ag-108m 127 27 5.58E-06 42.5 6897.5 1.51E-07 Sb-125 2.77 17 8.80E-06 67.2 6872.8 1.5E-07 Cs-134 2.06 45 3.35E-06 25.6 6914.4 1.51E-07 Cs-137 30 45 3.35E-06 25.6 6914.4 1.51E-07 Pm-147 2.62 95 1.59E-06 12.1 6927.9 1.51E-07 Eu-152 13.3 95 1.59E-06 12.1 6927.9 1.51E-07 Eu-154 8.8 95 1.59E-06 12.1 6927.9 1.51E-07 Eu-155 4.96 95 1.59E-06 12.1 6927.9 1.51E-07 Np-237 2140000 1

1.30E-04 990.8 5949.2 1.3E-07 Pu-238 87.7 174 8.70E-07 6.6 6933.4 1.51E-07 Pu-239 24100 174 8.70E-07 6.6 6933.4 1.51E-07 Pu-240 65400 174 8.70E-07 6.6 6933.4 1.51E-07 Pu-241 14.4 174 8.70E-07 6.6 6933.4 1.51E-07 Am-241 432 177 8.55E-07 6.5 6933.5 1.51E-07 Am-243 7380 177 8.55E-07 6.5 6933.5 1.51E-07 Cm-243 28.5 891 1.70E-07 1.3 6938.7 1.52E-07 Cm-244 18.11 891 1.70E-07 1.3 6938.7 1.52E-07 TSD 14-031 Revision 1 Page 19 of 25

15 3.2.4 Turbine Building The conceptual model assumes that the inventory is released instantly similar to the Containment Building. T he r esults of t his m odel a re pr esented i n T able 10. T he t able pr ovides t he maximum g roundwater concentration ( pCi/L), t he t able pr ovides t he a mount of r adioactivity (pCi) in solution, the amount sorbed to the solid material (pCi) and the peak concentration sorbed on the soil (pCi/g). The total inventory for each modeled nuclide in the Containment Building assuming a contamination level of 1 pCi/m2 on the floors and walls is 14679 pCi. The volume between the floor and the water table is 26135 m3, Table 1.

Table 10 Peak Groundwater and Soil Concentrations per unit source of 1 pCi/m2 in the Turbine Building.

Half-life (years)

Kd Peak Concentration (pCi/L)

Activity in Solution Activity Sorbed Peak Concentration (pCi/g)

H-3 12.4 0

2.25E-03 14679 0

0 C-14 5730 1.2 2.74E-04 1790.2 12888.8 3.29E-07 Fe-55 2.7 2857 1.31E-07 0.9 14678.1 3.74E-07 Ni-59 75000 62 6.02E-06 39.4 14639.6 3.73E-07 Co-60 5.27 223 1.68E-06 11 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 Nb-94 20300 45 8.29E-06 54.2 14624.8 3.73E-07 Tc-99 213000 0

2.25E-03 14679 0

0 Ag-108m 127 27 1.38E-05 90.1 14588.9 3.72E-07 Sb-125 2.77 17 2.18E-05 142.2 14536.8 3.71E-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 Pm-147 2.62 95 3.93E-06 25.7 14653.3 3.74E-07 Eu-152 13.3 95 3.93E-06 25.7 14653.3 3.74E-07 Eu-154 8.8 95 3.93E-06 25.7 14653.3 3.74E-07 Eu-155 4.96 95 3.93E-06 25.7 14653.3 3.74E-07 Np-237 2140000 1

3.21E-04 2097.1 12581.9 3.21E-07 Pu-238 87.7 174 2.15E-06 14.0 14665.0 3.74E-07 Pu-239 24100 174 2.15E-06 14.0 14665.0 3.74E-07 Pu-240 65400 174 2.15E-06 14.0 14665.0 3.74E-07 Pu-241 14.4 174 2.15E-06 14.0 14665.0 3.74E-07 Am-241 432 177 2.11E-06 13.8 14665.2 3.74E-07 Am-243 7380 177 2.11E-06 13.8 14665.2 3.74E-07 Cm-243 28.5 891 4.19E-07 2.7 14676.3 3.74E-07 Cm-244 18.11 891 4.19E-07 2.7 14676.3 3.74E-07 TSD 14-031 Revision 1 Page 20 of 25

16 3.2.5 Spent Fuel Building The conceptual model assumes that the inventory is released through diffusion through a one-half i nch uni formly c ontaminated z one on t he f loor, s imilar t o t he A uxiliary Building. The results of this model are presented in Table 11. The table provides the maximum groundwater concentration ( pCi/L), t he t able pr ovides t he a mount of r adioactivity (pCi) i n s olution, t he amount sorbed to the solid material (pCi) and the peak concentration sorbed on the soil (pCi/g).

The total inventory was 7.80E-10 Ci (780 pCi) for each nuclide modeled. The volume between the floor and the water table is 208 m3, Table 1.

Table 11 Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2 in the Spent Fuel Building Half-life Kd Diffusion Coefficient Peak Concentration Peak Radioactivity in Solution Peak Radioactivity Sorbed Peak Sorbed Concentration Nuclide (years)

(ml/g)

(cm2/s) pCi/L pCi pCi pCi/g H-3 12.4 0

5.5E-07 1.49E-02 774.8 0

0 C-14 5730 1.2 3.0E-09 1.83E-03 95.2 685.2 2.2E-06 Fe-55 2.7 2857 3.0E-09 2.04E-07 0.011 181.8 5.83E-07 Ni-59 75000 62 1.1E-09 4.02E-05 2.1 777.6 2.49E-06 Co-60 5.27 223 4.1E-11 4.25E-07 0.022 30 9.48E-08 Ni-63 96 62 1.1E-09 3.13E-05 1.6 605 1.94E-06 Sr-90 29.1 2.3 5.2E-10 3.21E-04 16.7 230.3 7.38E-07 Nb-94 20300 45 3.0E-09 5.53E-05 2.9 776.4 2.49E-06 Tc-99 213000 0

5.5E-07 1.50E-02 780 0.0 0

Ag-108m 127 27 3.0E-09 8.56E-05 4.5 721.1 2.31E-06 Sb-125 2.77 17 3.0E-09 3.41E-05 1.8 180.9 5.8E-07 Cs-134 2.06 45 3.0E-09 1.13E-05 0.6 158.7 5.09E-07 Cs-137 30 45 3.0E-09 4.07E-05 2.1 571.4 1.83E-06 Pm-147 2.62 95 3.0E-09 6.03E-06 0.3 178.7 5.73E-07 Eu-152 13.3 95 5.0E-11 1.75E-06 0.09 51.9 1.66E-07 Eu-154 8.8 95 5.0E-11 1.37E-06 0.07 40.6 1.3E-07 Eu-155 4.96 95 5.0E-11 1.04E-06 0.05 30.8 9.88E-08 Np-237 2140000 1

3.0E-09 2.14E-03 111.3 667.7 2.14E-06 Pu-238 87.7 174 3.0E-09 1.28E-05 0.67 694.9 2.23E-06 Pu-239 24100 174 3.0E-09 1.43E-05 0.74 776.3 2.49E-06 Pu-240 65400 174 3.0E-09 1.43E-05 0.74 776.3 2.49E-06 Pu-241 14.4 174 3.0E-09 7.83E-06 0.41 425.1 1.36E-06 Am-241 432 177 3.0E-09 1.38E-05 0.72 762.1 2.44E-06 Am-243 7380 177 3.0E-09 1.41E-05 0.73 778.7 2.5E-06 Cm-243 28.5 889 3.0E-09 2.03E-06 0.11 564.3 1.81E-06 Cm-244 18.11 889 3.0E-09 1.70E-06 0.09 472.6 1.51E-06 TSD 14-031 Revision 1 Page 21 of 25

17 3.2.6 Waste Water Treatment Facility The conceptual model assumes that the inventory is released instantly to the water, similar to the Containment Building. The results of this model are presented in Table 12. The table provides the maximum groundwater concentration (pCi/L), the table provides the amount of radioactivity (pCi) in solution, the amount sorbed to the solid material (pCi) and the peak concentration sorbed on t he s oil (pCi/g). T he t otal i nventory for each m odeled nuclide i n t he WWTF assuming a contamination level of 1 pC i/m2 on the floors and walls is 1124 pCi. T he volume between the floor and the water table is 144 m3, Table 1.

Table 12 Peak Groundwater and Soil Concentrations per unit source of 1 pCi/m2 in the Crib House/Forebay Building Half-life (years)

Kd Peak Concentration (pCi/L Activity in Solution Activity Sorbed Peak Concentration (pCi/g)

H-3 12.4 0

3.13E-02 1126 0

0 C-14 5730 1.2 3.82E-03 137.5 990.3 4.58E-06 Fe-55 2.7 2857 1.82E-06 0.1 1124.9 5.21E-06 Ni-59 75000 62 8.40E-05 3.0 1124.8 5.21E-06 Co-60 5.27 223 2.34E-05 0.8 1125.5 5.21E-06 Ni-63 96 62 8.40E-05 3.0 1124.8 5.21E-06 Sr-90 29.1 2.3 2.12E-03 76.2 1051.4 4.87E-06 Nb-94 20300 45 1.16E-04 4.2 1123.7 5.2E-06 Tc-99 213000 0

3.13E-02 1128 0

0 Ag-108m 127 27 1.92E-04 6.9 1120.9 5.19E-06 Sb-125 2.77 17 3.03E-04 10.9 1114.1 5.16E-06 Cs-134 2.06 45 1.16E-04 4.2 1123.4 5.2E-06 Cs-137 30 45 1.16E-04 4.2 1123.4 5.2E-06 Pm-147 2.62 95 5.48E-05 2.0 1125.3 5.21E-06 Eu-152 13.3 95 5.48E-05 2.0 1125.3 5.21E-06 Eu-154 8.8 95 5.48E-05 2.0 1125.3 5.21E-06 Eu-155 4.96 95 5.48E-05 2.0 1125.3 5.21E-06 Np-237 2140000 1

4.48E-03 161.1 966.7 4.48E-06 Pu-238 87.7 174 3.00E-05 1.1 1126.8 5.22E-06 Pu-239 24100 174 3.00E-05 1.1 1126.8 5.22E-06 Pu-240 65400 174 3.00E-05 1.1 1126.8 5.22E-06 Pu-241 14.4 174 3.00E-05 1.1 1126.8 5.22E-06 Am-241 432 177 2.95E-05 1.1 1126.8 5.22E-06 Am-243 7380 177 2.95E-05 1.1 1126.8 5.22E-06 Cm-243 28.5 891 5.85E-06 0.2 1124.9 5.21E-06 Cm-244 18.11 891 5.85E-06 0.2 1124.9 5.21E-06 TSD 14-031 Revision 1 Page 22 of 25

18 The WWTF has the hi ghest predicted concentrations of any building. This is because of the limited volume of water available for mixing. The ratio of the modeled inventory in the WWTF to the volume is 7.8 pCi/m3. This ratio in other buildings ranges from 3.75 pCi/m3 in the Spent Fuel Building down to 0.23 pCi/m3 in both the Auxiliary Building and the Crib House/Forebay Building. The predicted concentrations are consistent with the ratio of the modeled inventory to volume. The actual concentrations will be determined based on characterization data.

4.0 Conclusions A screening model for predicting peak groundwater and soil radionuclide concentrations at the Zion Nuclear Power Station after decommissioning has been developed. Values for each of the six underground structures that will remain after decommissioning are provided. Two structures, the Auxiliary Building and the Spent Fuel Building used a diffusion controlled release from the concrete as the conceptual model. The other four buildings, Containment, Crib House/Forebay, Turbine and the WWTF, assumed instant release of the entire inventory into the water/backfill region between the building floor and the water table. The choice of release model was based on existing data and process knowledge. The approach uses the DUST-MS simulation model to calculate the release and peak concentrations. The analysis is based on a unit source term of 1 pCi/m2 on the entire floor and walls of for each building. Conservative assumptions based on existing d ata w ere u sed i n t he s creening m odel for selecting p arameters t hat i mpact release (Diffusion coefficient) and groundwater concentration (Kd, porosity, bulk density). T he results can be c ombined w ith characterization da ta t o de termine pe ak groundwater dos e f or a ll t he nuclides and screen out those that are not significant contributors to dose.

TSD 14-031 Revision 1 Page 23 of 25

19 5.0 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.

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. 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.

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, 2001]

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., "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.

Szanto, Zs, Svingor, M. Molnir, L. Palcsu, I. Futo, Z. Szucs. "Diffusion of 3H, 99Tc, 125I, 36Cl, and 85Sr in granite, concrete and bentonite," Journal of Radioanalytical and Nuclear Chemistry, Vol. 252, No. 1 (2002) 133-138.

TSD 14-031 Revision 1 Page 24 of 25

20 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.

TSD 14-031 Revision 1 Page 25 of 25

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