ML17362A282

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Part 02 Ssar (Rev. 1) - Part 2 - Ssar - Chapter 2 - Site Characteristics - Section 2.4.13 - Accidental Releases of Liquid Effluents to Ground and Surface Waters
ML17362A282
Person / Time
Site: Clinch River
Issue date: 12/15/2017
From: James Shea
Tennessee Valley Authority
To:
Office of New Reactors
Fetter A
References
TVACLINCHRIVERESP, TVACLINCHRIVERESP.SUBMISSION.3, CRN.PART02, CRN.PART02.1, +reviewed
Download: ML17362A282 (28)
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Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report SUBSECTION 2.4.13 TABLE OF CONTENTS Section Title Page 2.4.13 Accidental Releases of Liquid Effluents to Ground and Surface Waters ......................................................................... 2.4.13-1 2.4.13.1 Accident Source .................................................... 2.4.13-1 2.4.13.2 Receptors ............................................................. 2.4.13-2 2.4.13.3 Primary Conceptual Model ................................... 2.4.13-2 2.4.13.4 Alternate Conceptual Model ................................. 2.4.13-2 2.4.13.5 Radionuclide Transport Analysis .......................... 2.4.13-3 2.4.13.6 Radionuclide Concentration in the Reservoir ....... 2.4.13-7 2.4.13.7 Dose Evaluation .................................................... 2.4.13-8 2.4.13.8 References ........................................................... 2.4.13-9 2.4.13-i Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report SUBSECTION 2.4.13 LIST OF TABLES Number Title 2.4.13-1 Liquid Release Source Terms vs Decay Time 2.4.13-2 Radionuclide Release Concentrations (C0) 2.4.13-3 Computation of Aquifer Bulk Density 2.4.13-4 Distribution Coefficient (Kd) Test Results and Summary Statistics 2.4.13-5 Transport/Dilution Analysis Parameters and Results 2.4.13-ii Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report SUBSECTION 2.4.13 LIST OF FIGURES Number Title 2.4.13-1 Minimum Distance from the Power Block Area to the Clinch River Arm of the Watts Bar Reservoir 2.4.13-2 Conceptual Model for Radionuclide Transport 2.4.13-3 Alternate Conceptual Model for Radionuclide Transport 2.4.13-iii Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report 2.4.13 Accidental Releases of Liquid Effluents to Ground and Surface Waters The purpose of this section is to provide a conservative analysis of a postulated, accidental release of radioactive liquid effluents to the environment at the Clinch River Nuclear (CRN) Site (Figure 2.4.13-1), in accordance with the guidance provided in Branch Technical Position (BTP) 11-6, Postulated Radioactive Releases Due to Liquid-Containing Tank Failures; DC/COL-ISG-013, Assessing the Radiological Consequences of Accidental Releases of Radioactive Materials from Liquid Waste Tanks for Combined License Applications (Reference 2.4.13-1); and DC/COL-ISG-014, Assessing the Radiological Consequences of Accidental Releases of Radioactive Materials from Liquid Waste Tanks in Ground and Surface Waters for Combined License Applications (Reference 2.4.13-2). The accident scenario is described and the methodology used to evaluate radionuclide transport is presented, along with potential contamination migration pathways to water users. The radionuclide concentrations and associated doses to which a water user might be exposed are compared against regulatory limits.

2.4.13.1 Accident Source It is postulated that a liquid radwaste tank outside of containment or outdoors ruptures with its contents released to the environment. The maximum tank volume with liquid radioactive waste is 10,000 gallons. The initial postulated radionuclide inventory of the spill is shown on Table 2.0-5.

A simplified release scenario is assumed, in which 80 percent of the tank volume is transferred instantaneously to the groundwater at a point within the power block area, and no credit is taken for the time that radionuclides may take (and the associated radioactive decay) to travel from the Liquid Waste Management System (LWMS) tank to the saturated zone. Analyzing a release of 80 percent of the tank volume is based on the guidance in DC/COL-ISG-013 (Reference 2.4.13-1) which states The radionuclide inventory for the tank and its components assumed to fail should be based on a conservative estimate of 80 percent capacity of that tank and its components. It is anticipated that a postulated radionuclide release will be mixed with groundwater of the CRN Site and will travel toward the Clinch River arm of the Watts Bar Reservoir.

Each of the four reactor vendors provided source term information for the accidental release of radioactive liquids, consistent with the guidance in NEI 10-01 (Reference 2.4.13-19). The source terms of the four SMR vendors were evaluated. The source term concentration for one vendor was found to be more conservative. On this basis, the source term and tank volume (10,000 gallons) associated with this vendor was adopted as the surrogate plant values.

However, based upon the large amount of conservatism included in the surrogate plant values as compared to the other SMR designs, a lower activity was used for Zr-95 and Nb-95. In addition, the surrogate plant values were compared to the PSEG ESPA source term values (Reference 2.4.13-20) to assess the reasonableness of the surrogate plant source terms. This comparison concluded that the surrogate plant source terms are conservative when compared to source terms for large light water reactors and are considered to be reasonable for use.

DC/COL-ISG-013 (Reference 2.4.13-1) also indicates that an applicant may take credit for mitigating design features provided they demonstrate that such features are durable and passive and that the receiving system has the storage capacity to hold the expected volume of liquid wastes. As a reactor technology has not been selected, no mitigation design features are considered in this analysis. Mitigation features included in the design of the reactor technology selected are addressed in the Combined License Application.

2.4.13-1 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report 2.4.13.2 Receptors NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 2.4.13, and BTP 11-6 require the consideration of radiation exposure to members of the public at points beyond the site boundary where the Applicant has no administrative control. Radiation doses are then calculated based on various consumption pathways on an annual basis as defined in Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I. The nearest site boundary beyond which the Tennessee Valley Authority (TVA) has no administrative control is the right bank (looking downstream) of the Clinch River arm of Watts Bar Reservoir (herein, referred to as the Reservoir).

The nearest surface water intake is the Oak Ridge Bear Creek Plant, located downstream of the CRN Site as shown on Table 2.4.1-1 and Figure 2.4.1-1 (Location No. 45). This plant is also known as the City of Oak Ridges West End Water Treatment Plant (WTP) and the K-25 Water Treatment Plant. The Oak Ridge Bear Creek Plant ceased water production on September 30, 2014, and Oak Ridge Utilities which now owns the facility has no plans to resume production at the site. Further downstream of the CRN Site, the closest surface water intakes in the Watts Bar Reservoir are near Kingston, TN. The Kingston Fossil Plant uses the Emory River/Watts Bar Reservoir as a source of thermoelectric cooling water, and the Kingston Water System uses the Tennessee River/Watts Bar Reservoir for public drinking water supply.

2.4.13.3 Primary Conceptual Model Figure 2.4.13-2 illustrates the primary conceptual model used to evaluate, for the CRN Site, an accidental liquid release of effluent to groundwater or to surface water via the groundwater pathway. The geology of the CRN Site consists of regolith composed of fill, clayey soils, and saprolite underlain by fractured bedrock, with fracture frequency decreasing with depth. As discussed in Subsection 2.5.1.2.5.1.2, solution widening of fractures has occurred, resulting in open and clay-filled cavities. Groundwater flow occurs primarily in the fractures and is approximately bounded within the upper 100 feet (ft) of subsurface material below ground surface as discussed in Subsection 2.4.12.1.3. Groundwater flow is toward the Reservoir.

Following release, it is assumed that radionuclides travel through the construction fill material and the shallow, pervasively fractured bedrock before reaching the Reservoir (Figure 2.4.13-2).

The travel distance is taken to be the minimum distance between the power block area and the Reservoir, which is a distance of 1400 ft. The shortest distance provides a conservative estimate, as less travel time allows for less decay of radionuclides.

During saturated zone transport, radionuclide concentrations of the liquid effluent released to the groundwater would be reduced by the processes of sorption, dispersion, and radioactive decay.

The key elements embodied in the conceptual model are described and discussed in Subsection 2.4.13.5.

2.4.13.4 Alternate Conceptual Model An alternate groundwater pathway involves groundwater discharge to the surface, via springs and seeps to onsite surface drainages and surface water discharge into the Clinch River arm of the Watts Bar Reservoir during wet periods (Figure 2.4.13-3). This scenario would be consistent with the conceptual model of groundwater flow in the Valley and Ridge Physiographic Province as discussed in Subsection 2.4.12.1.2. However, this alternate groundwater pathway is less conservative than the primary conceptual model, as additional surface water dilution from wet period runoff along with dilution in the Reservoir would lower radionuclide concentrations below those of the primary conceptual model. Thus, no further evaluation was undertaken to determine 2.4.13-2 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report radionuclide concentrations in the Reservoir based on the alternative conceptual model for radioactive release. As discussed in Subsection 2.4.12.3.1, it is very unlikely that there is shallow groundwater flow underneath the Clinch River arm of the Watts Bar Reservoir and exposure to water users on the opposite side of the Reservoir.

2.4.13.5 Radionuclide Transport Analysis A radionuclide transport analysis has been conducted to estimate the radionuclide concentrations that might result in exposure to existing and future water users based on an instantaneous release of the radioactive liquid from a hypothetical storage device. Analysis of liquid effluent release commences with a screening model, using demonstratively conservative assumptions and coefficients. Radionuclide concentrations resulting from the screening analysis are then compared to the effluent concentration limits (ECLs) identified in 10 CFR 20, Appendix B, Table 2, Column 2, to determine acceptability. Further analysis, using more realistic modeling techniques, is conducted for the radionuclides identified in the screening analysis.

2.4.13.5.1 Radionuclide Transport in Groundwater and Surface Water The effects of transport in groundwater and dilution in the Reservoir are modeled using an analytical approach outlined in Reference 2.4.13-3 for calculating the discharge rate (flux) of a radionuclide entering a surface water body that has intercepted the aquifer containing the transported material following an instantaneous release. The flux in this case is given by Equation 4.41 of Reference 2.4.13-3. Multiplying this equation by VT/Q gives the following expression for the dilution factor:





Equation 2.4.13-1 Where:

DL = dilution factor = C0/C C0 = source concentration C = concentration in the intercepting waterbody Dx = longitudinal dispersion coefficient [L2/T]

Dx = L

  • U Rd = retardation coefficient Q = flow rate of the Reservoir [L3/T]

L = longitudinal dispersivity of the aquifer [L]

VT = volume of release [L3]

U = pore velocity of groundwater [L/T]

= decay constant [T-1]

x = distance [L]

t = time [T]

Equation 2.4.13-1 allows calculation of DL as a function of time and can be solved by trial to find the minimum dilution factor, which yields the maximum concentration in the Reservoir. As 2.4.13-3 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report discussed in Reference 2.4.13-21, this equation accounts for the processes of advection, dispersion, sorption (retardation), and radioactive decay during groundwater transport, as well as dilution due to mixing with flows (Q) of the Reservoir.

2.4.13.5.2 Estimation of Initial Concentrations (C0)

Application of dilution factors obtained from Equation 2.4.13-1 requires an initial concentration (C0) for each radionuclide of concern. This is obtained by dividing the peak activity of the radionuclide by the tank volume. However, some daughter products (e.g., Pu-239) will arise after the release that are not included in the source term (Table 2.0-5), and others listed in the inventory will increase in activity as daughter products of a decay chain. Peak activity was determined by simulating decay of the source term inventory at various time steps ranging from 0 to 50 years for relevant radionuclides to capture the maximum activity of daughter products (Table 2.4.13-1).The maximum activity for each radionuclide was selected and divided by the release volume to arrive at C0 values for these radionuclides (Table 2.4.13-2).

Note that this approach overestimates the total activity released, as parent radionuclide activities are not decreased while daughter product activities are increased. With the exceptions of Nb-93m and U-235, all radionuclides showed peak activity well within the timescale of 50 years.

Nb-93m and U-235 do not reach peak activity within 50 years. Nb-93m was removed from further consideration because its parent radionuclide, Zr-93, is very long lived (half-life over 1 million years) and, therefore, significant concentrations of Nb-93m, which is relatively short-lived with a half-life of about 14 years, will not accumulate. As for U-235, which arises from the decay chain Np-239 Pu-239 U-235m U-235, near complete decay of Np-239 to U-235 would not occur for thousands of years, as the decay chain includes Pu-239, which is long-lived (half-life of about 24,000 years). However, computing the activity of U-235 that would result from the complete decay of the activity of its parent radionuclide, Np-239, gives an activity about 2.5 x 10-5 Curies (Ci), 700 times the value used in this analysis (3.56 x 10-8 Ci). This assumes each atom of Np-239 becomes an atom of U-235 and no decay occurs thereafter, and uses the fact that activity (in units of Ci) is equal to the product of the number of atoms times the radioactive decay constant (in units of s-1). Given that the results presented below show predicted exposure concentrations of U-235 on the order of 10-10 times the ECL, the initial activity value used for the analysis is not expected to have any significant effect on the subsequent dose analysis, as multiplying the concentration by 700 would still yield a value less than 10-6 of the ECL.

2.4.13.5.3 Input Parameters The input parameter values to determine the dilution factor, DL, as defined in Equation 2.4.13-1, are as follows:

2.4.13.5.3.1 Tank Volume and Volume Released (VT)

The maximum tank volume with liquid radioactive waste is 10,000 gallons. It is assumed that 80 percent of the tank volume is transferred instantaneously to the groundwater at a point within the power block area. Thus, 8000 gallons (1069.4 cubic ft) is used as input for VT in Equation 2.4.13-1.

2.4.13.5.3.2 Groundwater Pore Velocity (U) and Groundwater Travel Time The magnitude of groundwater pore velocity, U, is computed using the relation (Equation 4.8 of Reference 2.4.13-3):

2.4.13-4 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report K dH U = Equation 2.4.13-2 n e dx Where:

K = hydraulic conductivity [L/T]

ne = effective porosity (unitless) dH/dx = hydraulic gradient [L/L]

A hydraulic conductivity of 2.6 ft/day is used based on the results of the onsite aquifer pumping test analysis as discussed in Subsection 2.4.12.2.4.1. This is the computed hydraulic conductivity for the observation well oriented along the predominant groundwater flow path (parallel to strike).

The effective porosity was assigned a value of 0.0467 based on testing performed at the Oak Ridge Reservation (References 2.4.13-4 and 2.4.13-5) as discussed in Subsection 2.4.12.2.4.1.

The selected value is the average of 90 tests conducted using immersion-saturation porosimetry.

Based on site water level measurements, as discussed in Subsection 2.4.12.2.2, the horizontal hydraulic gradient at the CRN Site ranges from 0.03 to 0.11. The mean, 0.07, is used as the representative horizontal hydraulic gradient.

The resulting groundwater pore velocity (U), computed using Equation 2.4.13-2, is about 3.9 ft/day. The distance from the power block area to the Reservoir is approximately 1400 ft.

Thus, the groundwater travel time from the edge of the power block area to the Reservoir is estimated to be about 359 days (1400 ft/[3.9 ft/day]).

2.4.13.5.3.3 River Flow Rate (Q)

Outflow data for the Melton Hill Reservoir were used to assess the volumetric flow rate of the Reservoir near the CRN Site. Melton Hill Dam is located approximately five river miles upstream of the CRN Site. Daily average flow data were available from August 1962 to October 2013. This time range includes additional zero flow data associated with the early period of record before Melton Hill Dam was closed and filling of the reservoir was underway. The following statistics were calculated for this time period:

1. Daily average outflow rates range from 0 to nearly 35,000 cubic feet per second (cfs).
2. Zero flow was recorded for about 3.7 percent of the days in the period of record.
3. Daily average flow rate over the entire period of record is 4876 cfs.
4. Annual averages (based on calendar year) range from 2005 to 8071 cfs.
5. Lowest average flow rate over a continuous 365-day period was about 1760 cfs, which occurred from December 12, 2007 to December 10, 2008.

Additionally, TVA conducted its own analysis and determined the average weekly discharge from Melton Hill Dam over its lifetime to be approximately 4800 cfs with a maximum weekly discharge of approximately 25,450 cfs. TVA also analyzed expected flow frequency from Melton Hill Dam 2.4.13-5 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report based on 100 years of reservoir and system simulation conducted for the development of reservoir operating policy and determined a minimum flow requirement from Melton Hill Dam to be 400 cfs average daily flow. This minimum daily average release can be met, and has in the past been met, by operating the hydropower generating units for a period of only one hour per day. This can result in periods, potentially lasting up to 46 hr, where there are no releases from Melton Hill Dam. However, events during which there is no release from Melton Hill Dam for periods in excess of 36 hr are extremely rare. A bypass, which can produce a continuous flow rate of 400 cfs even when the hydropower generating units are not operating, will be installed at the dam.

Based on Equation 2.4.13-1, increase in flow (Q) in the Reservoir results in increase of dilution factor (DL), which in turn results in decrease in radionuclide concentrations in the Reservoir (C).

Taking a conservative approach (i.e., least dilution and maximum radionuclide concentration in the Reservoir), the minimum flow of 400 cfs average daily flow was used as input for Q in Equation 2.4.13-1. As noted in Reference 2.4.13-3, contaminated groundwater would enter the surface water as a diffuse patch as a result of source geometry (e.g., a pool of liquid on the ground surface) and dispersion processes which tend to spread the plume in all directions during transport, promoting mixing of contaminated groundwater seepage with the flow in the Reservoir.

The dilution flow rate of 400 cfs is selected to represent the near-field dilution (i.e., dilution at the interface of groundwater and surface water interaction) which results as groundwater enters the Reservoir via seepage through the riverbed prior to being available to a receptor.

2.4.13.5.3.4 Aquifer Bulk Density (b)

The aquifer bulk density is estimated based on the estimates of the geotechnical engineering properties. Table 2.4.13-3 presents estimates of the bulk density for each geologic formation at the CRN Site using relationships between saturated density, water content, and grain density (Reference 2.4.13-6). As shown in Table 2.4.13-4, measurements from the rock strata at the CRN Site, the primary media for groundwater transport and radionuclide travel, indicated a bulk density of about 2.7 g/cm3. However, the lowest value computed, 1.4 g/cm3 (for the existing fill and overlying soils), was selected to produce the lowest retardation coefficient (Equation 2.4.13-4) which is conservative for transport analysis.

2.4.13.5.3.5 Aquifer Longitudinal Dispersivity (L)

The longitudinal dispersivity was estimated using the relation between dispersivity and transport distance scale given in Equation 14b of Reference 2.4.13-7. This equation was chosen because it provides a higher weight to measured data points that have high reliability in Reference 2.4.13-7.

 Equation 2.4.13-3 The length scale, x, in the above equation is in meters and L is also in meters. The length scale used in the above equation is 426.7 m (1400 ft) for the transport distance from the edge of power block area to the edge of the Reservoir. The estimated L is 8.57 m (28.1 ft).

2.4.13.5.3.6 Radioactive Half-Life and Decay Constant Radioactive half-lives for the radionuclides under consideration were obtained from References 2.4.13-8, 2.4.13-9, and 2.4.13-10. Radionuclide half-lives were used to calculate the decay constant for each radionuclide, which are given in Table 2.4.13-2.

2.4.13-6 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report 2.4.13.5.3.7 Distribution Coefficients (Kd)

Distribution coefficients, Kd, were applied to selected radionuclides based on laboratory testing carried out using samples from the CRN Site as shown on Table 2.4.13-4. Kd is used in the computation of the retardation coefficient, Rd, using the relation:

Equation 2.4.13-4 Where b is the bulk density and ne is the effective porosity of the aquifer. Non-zero Kd values were used for elements for which site-specific laboratory measurements were available. The geometric mean of the laboratory-derived Kd values was assigned as the representative value for each element. Kd test result summary statistics and analysis notes are presented in Table 2.4.13-4. No site-specific Kd value was measured for the Yttrium (Y) series of radioisotopes, such as Y-91. However, Y chemistry is similar to a lanthanide and is often associated in the environment with lanthanides such as cerium (Ce) (Reference 2.4.13-11). The site-specific geometric mean Kd value for Ce is 54 mL/g and thus the likely Kd value for Y is also 54 mL/g. Table 2.4.13-5 presents the resulting Rd values as well as the associated solute travel times (used as t in Equation 2.4.13-1) which are computed as the product of the groundwater travel time and Rd.

2.4.13.6 Radionuclide Concentration in the Reservoir Minimum dilution factors and associated maximum concentrations in the Reservoir are provided in Table 2.4.13-5. For screening purposes, dilution factors were initially calculated without considering the effects of sorption (using Equation 2.4.13-1 with Rd = 1). The resulting concentrations in the Reservoir were compared against the ECLs. This resulted in several exceedances of ECLs.

Further analysis accounting for sorption of the radioisotopes listed in Subsection 2.4.13.5.3.7 (Table 2.4.13-4) was performed considering the effects of sorption/retardation. This resulted in a decrease of the estimated concentrations in the Reservoir to below the ECL for all isotopes (Table 2.4.13-5). These concentrations were carried forward for dose evaluation, as described in the Subsection 2.4.13.7.

The resultant concentrations in the Reservoir in Table 2.4.13-5 (with sorption) are used to develop the annual dose estimates and the following conservatisms have been included in their determination.

The radionuclide release was assumed to enter the groundwater instantaneously with no consideration of any containment barrier, radionuclide decay, or onsite surface water dilution as a result of wet periods, and all radionuclide concentrations (including daughter products) are assumed to be coincidentally at their peak. This is conservative because it overestimates the concentrations and the annual dose for many of the radionuclides of interest.

The flow rate in the Reservoir is assumed to be 400 cfs, which represents the minimum release requirement for the upstream Melton Hill Dam per its reservoir operating policy. The value of 400 cfs is 4.4 times lower than the minimum daily average flow rate over a continuous 365-day period (1760 cfs) and 12.2 times lower than the daily average flow rate (4876 cfs). Also, the 400 cfs flow rate assumes no tributary or groundwater inflows between the Melton Hill Dam and the CRN Site, which are reasonably expected to occur and which would increase flow downstream of the dam. This lower assumed flow rate results in higher radionuclide concentrations at the receptor and consequently higher doses.

2.4.13-7 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report The distribution coefficient for radionuclides, for which no site-specific distribution coefficients are available (e.g., Nb-95, Sb-124), is assumed to be zero, even though non-zero values are reported in literature. A larger retardation factor (associated with non-zero distribution coefficients) would result in longer radionuclide travel time, more radioactive decay, and lower concentrations in the Reservoir. For example, Reference 2.4.13-18 reports distribution coefficient values for Nb ranging from 80 to 100 cm3/g, based on batch and column experiments conducted with crushed rock. If similar values were used in the analysis, the calculated Nb-95 concentration in the Reservoir would be essentially zero, and the dose to the receptor would be greatly reduced.

Finally, the travel time for the release is that associated with the shortest path between the power block area and the Reservoir, which limits radioactive decay and results in higher concentrations in the Reservoir and higher doses.

2.4.13.7 Dose Evaluation In addition to meeting the 10 CFR 20, Appendix B, ECLs, the dose due to the radionuclide concentrations in the Reservoir (Table 2.4.13-5) must also meet the 10 CFR 20.1301 dose limit for a member of the public. The Reservoir is a potential source of drinking water, as well as of aquatic foods and is also used for recreational activities. The LADTAP II computer program (Reference 2.4.13-12) is used to calculate dose associated with liquid radioactive waste effluent and was used in the evaluation of doses due to accidental release of radioactive liquids.

Whereas LADTAP II typically calculates total body and organ doses, the dose limit in 10 CFR 20.1301 is in terms of total effective dose equivalent (TEDE). As dose is directly proportional to the dose conversion factor (DCF), the dose calculated by LADTAP II for a given nuclide and exposure pathway was adjusted to TEDE by changing the DCFs applied (in the DCF library within the LADTAP II code).

The following exposure pathways were considered in evaluating dose:

Consumption of water from the Reservoir Consumption of fish and invertebrate from the Reservoir Consumption of vegetables, milk, and meat affected by irrigation water from the Reservoir Boating, swimming, and shoreline activities on the Reservoir The major inputs and assumptions are as follows:

No dilution is credited beyond the concentrations shown in Table 2.4.13-5 Transit time to dose receptors is assumed to be zero Irrigation rate is assumed to be 1 inch/week, which bounds the actual rate near the plant of 0.24 inch/week (Reference 2.4.13-13)

Consumption and usage rates are the default values for the maximally exposed individual from Regulatory Guide 1.109, Table E-5, while assuming that the time spent on boating and swimming is each the same as that for shoreline activities Exposure duration is assumed to be 1 year 2.4.13-8 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report TEDE dose conversion factors for ingestion are obtained from Federal Guidance Report 11 (Reference 2.4.13-14)

TEDE dose conversion factors for ground deposition and immersion are obtained from Federal Guidance Report 12 (Reference 2.4.13-15)

LADTAP II requires the source nuclide activities to be input in the units of Ci/yr. The concentrations in Table 2.4.13-5 are converted into release rates (Ci/yr) by multiplying by an arbitrary flow rate of 1 cfs, which is then entered as the liquid effluent discharge rate. LADTAP II divides the isotopic activity by the discharge rate to obtain the initial concentrations, causing the flow rate to cancel out. With the transit time to the dose receptor input as zero, LADTAP II calculates concentrations at the receptor that are identical to those in Table 2.4.13-5. The resulting total dose from all exposure pathways is 93 mrem TEDE to an adult, the age group receiving the maximum dose. This dose is within the 10 CFR 20.1301 limit of 100 mrem TEDE.

2.4.13.8 References 2.4.13-1. NRC, DC/COL-ISG-013, Assessing the Radiological Consequences of Accidental Releases of Radioactive Materials from Liquid Waste Tanks for Combined License Applications, January 2013.

2.4.13-2. NRC, DC/COL-ISG-014, Assessing the Radiological Consequences of Accidental Releases of Radioactive Materials from Liquid Waste Tanks in Ground and Surface Waters for Combined License Applications, January 2013.

2.4.13-3. NUREG/CR-3332, Radiological Assessment: A Textbook on Environmental Dose Analysis, pp. 4-1-4-53, September 1983.

2.4.13-4. Oak Ridge National Laboratory Environmental Services Division, ORNL/GWPO-021, Effective Porosity and Pore-Throat Sizes of Conasauga Group Mudrock: Application, Test and Evaluation of Petrophysical Techniques, April 1996.

2.4.13-5. Oak Ridge National Laboratory Environmental Services Division, ORNL/GWPO-026, Effective Porosity and Density of Carbonate Rocks (Maynardville Limestone and Copper Ridge Dolomite) within Bear Creek Valley on the Oak Ridge Reservation Based on Modern Petrophysical Techniques, February 1997.

2.4.13-6. Terzaghi, K., and R.B. Peck, Soil Mechanics in Engineering Practice, Wiley, 2d ed.,

1967.

2.4.13-7. Xu, M., and Y. Eckstein, Use of Weighted Least-Squares Method in Evaluation of the Relationship Between Dispersivity and Field Scale, Groundwater, Vol. 33, Issue 6, pp. 905-908, 1995.

2.4.13-8. NUREG/CR-5512, Residual Radioactive Contamination From Decommissioning, Vol. 1, October 1992.

2.4.13-9. International Commission on Radiation Protection, ICRP Publication 107, Annals ICRP, Vol. 38, Issue 3, Nuclear Decay Data for Dosimetric Calculations, 2008.

2.4.13-10. Firestone, R.B. et al., Update to the 8th edition of the Table of Isotopes, Booklet and updated CD-ROM, John Wiley & Sons, Inc., 1999.

2.4.13-9 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report 2.4.13-11. Strenge, D.L., and S.R. Peterson, Chemical data bases for the Multimedia Environmental Pollutant Assessment System (MEPAS): Ver. 1, Prepared by Pacific Northwest National Laboratory, December 1989.

2.4.13-12. NUREG/CR-4013, LADTAP II - Technical Reference and User Guide, April 1986.

2.4.13-13. TVA (River Operations and Renewables), Water Use in the Tennessee Valley for 2010 and Projected Use in 2035, July 2012.

2.4.13-14. Environmental Protection Agency, EPA 520/1-88-020, Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, 1988.

2.4.13-15. EPA, EPA-402-R-93-081, Federal Guidance Report No. 12, External Exposure to Radionuclides in Air, Water, and Soil, 1993.

2.4.13-16. Nehls, G.J., and G.G. Akland, Procedures for Handling Aerometric Data, Journal of the Air Pollution Control Association, Vol. 23, pp.180-184, 1973.

2.4.13-17. Moore, G.K., Quantification of Ground-Water Flow in Fractured Rock, Oak Ridge, Tennessee, Groundwater, Vol. 35, No. 3, May-June, 1997.

2.4.13-18. Anderson, K., Torstenfelt, B., and Rydberg, J., Leakage of Niobium-94 from an Underground Rock Repository, Department of Nuclear Chemistry, Chalmers University of Technology, Gteborg, Sweden, 1979. Available at http://www.iaea.org/inis/collection/NCLCollectionStore/_

Public/11/561/11561931.pdf 2.4.13-19. NEI 10-01, Industry Guidance for Developing a Plant Parameter Envelope in Support of an Early Site Permit, Revision 1, May 2012.

2.4.13-20. PSEG Power, LLC, Application for Early Site Permit for the PSEG Site, Revision 4, June 5, 2015.

2.4.13-21. Taylor, S.W. and H. Guha, Time and Magnitude of Peak Concentration of Reactive Groundwater Contaminants Discharged to a River, Groundwater, Volume 55, No.

1, January-February 2017, pps. 63-72.

2.4.13-10 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Table 2.4.13-1 (Sheet 1 of 4)

Liquid Release Source Terms vs Decay Time Activity (Ci) vs Decay Time 24 hr 168 hr 720 hr 8760 hr 26280 hr Isotope 0 hr 1 hr 1 day 100 hr 1 wk 1 mon 1 yr 3 yr H-3 1.25 x 102 1.25 x 102 1.25 x 102 1.25 x 102 1.25 x 102 1.24 x 102 1.18 x 102 1.06 x 102

-1 -1 -1 -1 -1 -1 -1 C-14 1.37 x 10 1.37 x 10 1.37 x 10 1.37 x 10 1.37 x 10 1.37 x 10 1.37 x 10 1.37 x 10-1 Na-24 7.68 x 101 7.33 x 101 2.54 x 101 7.63 x 10-1 3.32 x 10-2 2.92 x 10-13 0 0 P-32 1.95 x 101 1.95 x 10 1 1.86 x 10 1 1.59 x 10 1 1.39 x 10 1 4.55 x 10 0 3.97 x 10-7 1.64 x 10-22 3 3 3 3 3 3 -1 Cr-51 8.04 x 10 8.03 x 10 7.84 x 10 7.24 x 10 6.75 x 10 3.80 x 10 8.69 x 10 1.01 x 10-8 Mn-54 7.49 x 102 7.49 x 102 7.47 x 102 7.42 x 102 7.38 x 102 7.01 x 102 3.33 x 102 6.59 x 101 4 4 1 -8 -16 Mn-56 2.26 x 10 1.73 x 10 3.59 x 10 4.89 x 10 5.71 x 10 0 0 0 Fe-55 2.99 x 103 2.99 x 103 2.99 x 103 2.98 x 103 2.98 x 103 2.93 x 103 2.31 x 103 1.38 x 103 Fe-59 1.93 x 102 1.93 x 102 1.90 x 102 1.81 x 102 1.73 x 102 1.21 x 102 6.61 x 10-1 7.76 x 10-6 Co-58 1.21 x 103 1.21 x 103 1.20 x 103 1.16 x 103 1.13 x 103 9.02 x 102 3.40 x 101 2.68 x 10-2 Co-60 2.59 x 102 2.59 x 102 2.59 x 102 2.59 x 102 2.58 x 102 2.56 x 102 2.27 x 102 1.75 x 102 Ni-63 9.63 x 101 9.63 x 101 9.63 x 101 9.63 x 101 9.63 x 101 9.63 x 101 9.56 x 101 9.43 x 101 Cu-64 5.25 x 10-1 4.97 x 10-1 1.42 x 10-1 2.24 x 10-3 5.47 x 10-5 4.52 x 10-18 0 0 Zn-65 8.50 x 10-5 8.50 x 10-5 8.48 x 10-5 8.40 x 10-5 8.33 x 10-5 7.81 x 10-5 3.02 x 10-5 3.79 x 10-6 Rb-89 1.30 x 105 8.36 x 103 3.21 x 10-24 0 0 0 0 0 Sr-89 1.34 x 105 1.34 x 105 1.32 x 105 1.27 x 105 1.22 x 105 8.88 x 104 8.99 x 102 4.04 x 10-2 Sr-90 1.87 x 104 1.87 x 104 1.87 x 104 1.87 x 104 1.87 x 104 1.87 x 104 1.83 x 104 1.74 x 104 Sr-91 1.71 x 105 1.59 x 105 2.96 x 104 1.14 x 102 7.91 x 10-1 2.35 x 10-18 0 0 Sr-92 1.84 x 105 1.43 x 105 3.98 x 102 1.44 x 10-6 4.06 x 10-14 0 0 0 Y-90 1.94 x 104 1.94 x 104 1.92 x 104 1.89 x 104 1.88 x 104 1.87 x 104 1.83 x 104 1.74 x 104 Y-91m 0 5.30 x 104 1.86 x 104 7.45 x 101 5.12 x 10-1 3.33 x 10-18 0 0 Y-91 1.76 x 105 1.76 x 105 1.75 x 105 1.69 x 105 1.63 x 105 1.24 x 105 2.35 x 103 4.12 x 10-1 Y-92 1.86 x 105 1.82 x 105 5.88 x 103 2.49 x 10-3 4.13 x 10-9 0 0 0 5 1.96 x 105 4.18 x 104 2.51 x 102 2.59 x 100 1.92 x 10-16 0 0 Y-93 2.10 x 10 Zr-93 0 1.05 x 10-5 1.29 x 10-4 1.61 x 10-4 1.61 x 10-4 1.61 x 10-4 1.61 x 10-4 1.61 x 10-4 Zr-95 5.00 x 103 5.00 x 103 4.95 x 103 4.78 x 103 4.64 x 103 3.61 x 103 9.58 x 101 3.52 x 10-2 Nb-93m 0 2.82 x 10-11 1.05 x 10-8 7.53 x 10-8 1.36 x 10-7 6.27 x 10-7 7.60 x 10-6 2.17 x 10-5 Nb-95m 0 3.59 x 10-1 7.82 x 100 2.42 x 101 3.18 x 101 3.43 x 101 9.40 x 10-1 3.73 x 10-4 Nb-95 5.00 x 103 5.00 x 103 5.00 x 103 4.99 x 103 4.97 x 103 4.64 x 103 2.08 x 102 7.89 x 10-2 Mo-99 2.65 x 105 2.62 x 105 2.06 x 105 9.28 x 104 4.54 x 104 1.38 x 102 0 0 Tc-99m 2.35 x 105 2.34 x 10 5 1.96 x 10 5 8.89 x 10 4 4.35 x 10 4 1.38 x 102 0 0 Tc-99 0 9.95 x 10-5 2.20 x 10-3 6.53 x 10-3 8.35 x 10-3 1.01 x 10-2 1.01 x 10-2 1.01 x 10-2 Ru-103 2.24 x 105 2.24 x 105 2.20 x 105 2.08 x 105 1.98 x 105 1.32 x 105 3.62 x 102 9.48 x 10-4 4 4 4 4 4 4 4 Ru-106 8.63 x 10 8.63 x 10 8.61 x 10 8.56 x 10 8.52 x 10 8.15 x 10 4.33 x 10 1.09 x 104 Rh-103m 2.24 x 105 2.24 x 105 2.20 x 105 2.08 x 105 1.98 x 105 1.33 x 105 4.08 x 102 1.24 x 10-3 Rh-106 9.16 x 104 8.63 x 104 8.61 x 104 8.56 x 104 8.52 x 104 8.16 x 104 4.38 x 104 1.12 x 104 Ag-110m 4.31 x 102 4.31 x 102 4.30 x 102 4.26 x 102 4.23 x 102 3.97 x 102 1.58 x 102 2.13 x 101 Ag-110 0 6.46 x 100 6.45 x 100 6.39 x 100 6.34 x 100 5.96 x 100 2.42 x 100 3.32 x 10-1 2.4.13-11 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Table 2.4.13-1 (Sheet 2 of 4)

Liquid Release Source Terms vs Decay Time Activity (Ci) vs Decay Time 24 hr 168 hr 720 hr 8760 hr 26280 hr Isotope 0 hr 1 hr 1 day 100 hr 1 wk 1 mon 1 yr 3 yr Sb-124 1.18 x 102 1.18 x 102 1.17 x 102 1.13 x 102 1.09 x 102 8.36 x 101 1.77 x 100 3.97 x 10-4 Te-129m 7.21 x 103 7.20 x 103 7.06 x 103 6.62 x 103 6.24 x 103 3.88 x 103 3.81 x 100 1.06 x 10-6 3 3 3 3 3 0 Te-129 0 2.04 x 10 4.46 x 10 4.18 x 10 3.94 x 10 2.47 x 10 2.76 x 10 9.18 x 10-7 Te-131m 2.73 x 104 2.67 x 104 1.57 x 104 2.71 x 103 5.63 x 102 1.63 x 10-3 0 0 3 3 2 2 Te-131 0 4.80 x 10 3.51 x 10 6.18 x 10 1.28 x 10 4.68 x 10-4 0 0 Te-132 2.02 x 105 2.00 x 10 5 1.63 x 10 5 8.33 x 10 4 4.56 x 10 4 3.42 x 10 2 3.86 x 10 -29 0

I-129 5.89 x 10-3 5.89 x 10 -3 5.89 x 10 -3 5.89 x 10 -3 5.89 x 10 -3 5.90 x 10 -3 5.92 x 10 -3 5.92 x 10-3 3 3 2 0 -1 -15 I-130 2.42 x 10 2.29 x 10 6.30 x 10 8.90 x 10 1.97 x 10 7.19 x 10 0 0 I-131 1.42 x 105 1.42 x 105 1.32 x 105 1.02 x 105 8.03 x 104 1.11 x 104 3.19 x 10-9 0 I-132 2.07 x 105 2.05 x 10 5 1.68 x 10 5 8.59 x 10 4 4.70 x 10 4 3.78 x 10 2 2.20 x 10 -28 0

I-133 2.92 x 105 2.83 x 105 1.32 x 105 1.06 x 104 1.11 x 103 1.25 x 10-5 0 0 I-134 3.28 x 105 1.49 x 105 1.82 x 10-3 1.32 x 10-29 0 0 0 0 I-135 2.78 x 105 2.50 x 105 2.25 x 104 7.79 x 100 6.24 x 10-3 4.61 x 10-28 0 0 0

Xe-131m 0 3.82 x 10 8.61 x 101 2.89 x 102 3.98 x 102 3.32 x 102 1.67 x 10-6 3.59 x 10-25 Xe-133m 0 1.09 x 102 1.55 x 103 1.29 x 103 5.89 x 102 4.21 x 10-1 0 0 Xe-133 0 1.53 x 103 2.41 x 104 3.11 x 104 2.29 x 104 1.13 x 103 7.01 x 10-17 0 4

Xe-135m 0 3.73 x 10 3.69 x 103 1.42 x 100 1.12 x 10-3 2.82 x 10-28 0 0 Xe-135 0 1.83 x 104 5.93 x 104 3.46 x 102 2.05 x 100 1.16 x 10-18 0 0 Cs-134 3.01 x 104 3.01 x 104 3.01 x 104 3.00 x 104 2.99 x 104 2.93 x 104 2.15 x 104 1.10 x 104 Cs-135 0 2.47 x 10-7 4.45 x 10-5 7.10 x 10-5 7.11 x 10-5 7.11 x 10-5 7.11 x 10-5 7.11 x 10-5 Cs-136 1.00 x 104 9.98 x 103 9.48 x 103 8.01 x 103 6.89 x 103 2.02 x 103 3.55 x 10-5 4.46 x 10-22 Cs-137 2.45 x 104 2.45 x 104 2.45 x 104 2.45 x 104 2.45 x 104 2.45 x 104 2.39 x 104 2.29 x 104 Cs-138 2.71 x 105 7.45 x 104 9.41 x 10-9 0 0 0 0 0 Ba-136m 0 1.50 x 103 1.42 x 103 1.21 x 103 1.04 x 103 3.11 x 102 7.76 x 10-6 1.61 x 10-22 Ba-137m 0 2.32 x 104 2.32 x 104 2.32 x 104 2.32 x 104 2.31 x 104 2.27 x 104 2.17 x 104 Ba-140 2.50 x 105 2.49 x 105 2.37 x 105 2.00 x 105 1.71 x 105 4.92 x 104 6.43 x 10-4 4.24 x 10-21 La-140 2.58 x 105 2.58 x 105 2.53 x 105 2.24 x 105 1.95 x 105 5.67 x 104 9.48 x 10-4 1.04 x 10-20 Ce-141 2.36 x 105 2.36 x 105 2.31 x 105 2.16 x 105 2.03 x 105 1.25 x 105 9.96 x 101 1.77 x 10-5 Ce-144 2.02 x 105 2.02 x 105 2.02 x 105 2.00 x 105 1.99 x 105 1.88 x 105 8.30 x 104 1.40 x 104 Pr-143 2.15 x 105 2.15 x 105 2.04 x 105 1.74 x 105 1.51 x 105 4.66 x 104 1.78 x 10-3 1.21 x 10-19 Pr-144m 0 3.02 x 103 3.02 x 103 3.00 x 103 2.98 x 103 2.82 x 103 1.27 x 103 2.18 x 102 Pr-144 0 1.84 x 105 2.02 x 105 2.00 x 105 1.99 x 105 1.88 x 105 8.44 x 104 1.45 x 104 Nd-144 0 0 0 5.44 x 10-13 1.08 x 10-12 5.27 x 10-12 4.42 x 10-11 6.98 x 10-11 U-235m 0 3.28 x 100 1.78 x 10-1 5.21 x 10-1 6.37 x 10-1 7.24 x 10-1 7.25 x 10-1 7.24 x 10-1 U-235 0 2.34 x 10-13 6.94 x 10-13 4.03 x 10-12 8.54 x 10-12 5.26 x 10-11 7.07 x 10-10 2.13 x 10-9 Np-239 2.72 x 106 2.69 x 106 2.02 x 106 7.94 x 105 3.44 x 105 3.85 x 102 0 0 Pu-239 0 8.86 x 10-3 1.85 x 10-1 5.13 x 10-1 6.33 x 10-1 7.24 x 10-1 7.25 x 10-1 7.24 x 10-1 Total 8.10 x 106 7.77 x 106 5.66 x 106 3.55 x 106 2.76 x 106 1.45 x 106 3.68 x 105 1.43 x 105 2.4.13-12 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Table 2.4.13-1 (Sheet 3 of 4)

Liquid Release Source Terms vs Decay Time Activity (Ci) vs Decay Time Time of 43800 hr 87600 hr 131400 hr 219000 hr 306600 hr 438000 hr Peak Peak Isotope 5 yr 10 yr 15 yr 25 yr 35 yr 50 yr Activity (Ci) Activity H-3 9.44 x 101 7.13 x 101 5.39 x 101 3.07 x 101 1.75 x 101 7.55 x 100 1.25 x 102 0 hr C-14 1.37 x 10-1 1.37 x 10-1 1.37 x 10-1 1.37 x 10-1 1.36 x 10-1 1.36 x 10-1 1.37 x 10-1 0 hr Na-24 0 0 0 0 0 0 7.68 x 101 0 hr P-32 0 0 0 0 0 0 1.95 x 101 0 hr Cr-51 1.19 x 10-16 0 0 0 0 0 8.04 x 103 0 hr Mn-54 1.30 x 101 2.27 x 10 -1 3.95 x 10 -3 1.20 x 10 -6 3.63 x 10 -10 1.92 x 10 -15 7.49 x 10 2 0 hr Mn-56 0 0 0 0 0 0 2.26 x 104 0 hr Fe-55 8.21 x 102 2.25 x 102 6.18 x 101 4.66 x 100 3.51 x 10-1 7.25 x 10-3 2.99 x 103 0 hr Fe-59 9.10 x 10-11 4.30 x 10-23 0 0 0 0 1.93 x 102 0 hr Co-58 2.11 x 10-5 3.68 x 10 -13 6.41 x 10 -21 0 0 0 1.21 x 10 3 0 hr Co-60 1.34 x 102 6.96 x 101 3.61 x 101 9.69 x 100 2.60 x 100 3.62 x 10-1 2.59 x 102 0 hr Ni-63 9.30 x 101 8.99 x 101 8.68 x 101 8.10 x 101 7.56 x 101 6.81 x 101 9.63 x 101 0 hr Cu-64 0 0 0 0 0 0 5.25 x 10-1 0 hr

-7 Zn-65 4.77 x 10 2.68 x 10-9 1.50 x 10-11 4.73 x 10-16 1.49 x 10-20 2.63 x 10-27 8.50 x 10-5 0 hr Rb-89 0 0 0 0 0 0 1.30 x 105 0 hr Sr-89 1.82 x 10-6 2.47 x 10-17 3.34 x 10-28 0 0 0 1.34 x 105 0 hr 4

Sr-90 1.66 x 10 1.47 x 104 1.30 x 104 1.03 x 104 8.06 x 103 5.62 x 103 1.87 x 104 0 hr Sr-91 0 0 0 0 0 0 1.71 x 105 0 hr Sr-92 0 0 0 0 0 0 1.84 x 105 0 hr 4

Y-90 1.66 x 10 1.47 x 104 1.31 x 104 1.03 x 104 8.10 x 103 5.66 x 103 1.94 x 104 0 hr Y-91m 0 0 0 0 0 0 5.30 x 104 1 hr Y-91 7.24 x 10-5 2.96 x 10-14 1.21 x 10-23 0 0 0 1.76 x 105 0 hr 5

Y-92 0 0 0 0 0 0 1.86 x 10 0 hr Y-93 0 0 0 0 0 0 2.10 x 105 0 hr Zr-93 1.61 x 10-4 1.61 x 10-4 1.61 x 10-4 1.61 x 10-4 1.61 x 10-4 1.61 x 10-4 1.61 x 10-4 1 wk Zr-95 1.29 x 10-5 3.34 x 10-14 8.65 x 10-23 0 0 0 5.00 x 103 0 hr Nb-93m 3.45 x 10-5 6.12 x 10-5 8.19 x 10-5 1.10 x 10-4 1.28 x 10-4 1.41 x 10-4 1.41 x 10-4 50 yr Nb-95m 1.37 x 10-7 4.59 x 10-16 1.19 x 10-24 0 0 0 3.43 x 101 1 mon Nb-95 2.90 x 10-5 8.11 x 10-14 2.10 x 10-22 0 0 0 5.00 x 103 0 hr Mo-99 0 0 0 0 0 0 2.65 x 105 0 hr Tc-99m 0 0 0 0 0 0 2.35 x 105 0 hr Tc-99 1.01 x 10-2 1.01 x 10-2 1.01 x 10-2 1.01 x 10-2 1.01 x 10-2 1.01 x 10-2 1.01 x 10-2 1 yr Ru-103 2.48 x 10-9 2.74 x 10 -23 0 0 0 0 2.24 x 10 5 0 hr Ru-106 2.74 x 103 8.67 x 101 2.75 x 100 2.76 x 10-3 2.77 x 10-6 8.83 x 10-11 8.63 x 104 0 hr Rh-103m 3.23 x 10-9 5.57 x 10-23 0 0 0 0 2.24 x 105 0 hr Rh-106 2.81 x 103 9.30 x 101 2.95 x 100 3.18 x 10-3 3.19 x 10-6 1.09 x 10-10 9.16 x 104 0 hr Ag-110m 2.86 x 100 1.90 x 10-2 1.26 x 10-4 5.58 x 10-9 2.46 x 10-13 7.21x 10-20 4.31 x 102 0 hr Ag-110 4.47 x 10-2 3.16 x 10-4 2.10 x 10-6 1.03 x 10-10 4.54 x 10-15 1.48 x 10-21 6.46 x 100 1 hr 2.4.13-13 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Table 2.4.13-1 (Sheet 4 of 4)

Liquid Release Source Terms vs Decay Time Activity (Ci) vs Decay Time Time of 43800 hr 87600 hr 131400 hr 219000 hr 306600 hr 438000 hr Peak Peak Isotope 5 yr 10 yr 15 yr 25 yr 35 yr 50 yr Activity (Ci) Activity Sb-124 8.92 x 10-8 6.74 x 10-17 5.10 x 10-26 0 0 0 1.18 x 102 0 hr Te-129m 2.96 x 10-13 1.21 x 10-29 0 0 0 0 7.21 x 103 0 hr Te-129 2.56 x 10-13 1.78 x10 -29 0 0 0 0 4.46 x 10 3 1 day Te-131m 0 0 0 0 0 0 2.73 x 104 0 hr Te-131 0 0 0 0 0 0 4.80 x 103 1 hr Te-132 0 0 0 0 0 0 2.02 x 105 0 hr I-129 5.92 x 10-3 5.92 x 10 -3 5.92 x 10 -3 5.92 x 10 -3 5.92 x 10 -3 5.92 x 10 -3 5.92 x 10 -3 1 yr I-130 0 0 0 0 0 0 2.42 x 103 0 hr I-131 0 0 0 0 0 0 1.42 x 105 0 hr I-132 0 0 0 0 0 0 2.07 x 105 0 hr I-133 0 0 0 0 0 0 2.92 x 105 0 hr I-134 0 0 0 0 0 0 3.28 x 105 0 hr I-135 0 0 0 0 0 0 2.78 x 105 0 hr 2

Xe-131m 0 0 0 0 0 0 3.98 x 10 1 wk Xe-133m 0 0 0 0 0 0 1.55 x 103 1 day Xe-133 0 0 0 0 0 0 3.11 x 104 100 hr Xe-135m 0 0 0 0 0 0 3.73 x 104 1 hr Xe-135 0 0 0 0 0 0 5.93 x 104 1 day Cs-134 5.61 x 103 1.04 x 103 1.94 x 102 6.74 x 100 2.34 x 10-1 1.51 x 10-3 3.01 x 104 0 hr

-5 Cs-135 7.11 x 10 7.11 x 10-5 7.11 x 10-5 7.11 x 10-5 7.11 x 10-5 7.11 x 10-5 7.11 x 10-5 1 wk Cs-136 0 0 0 0 0 0 1.00 x 104 0 hr Cs-137 2.18 x 104 1.95 x 104 1.74 x 104 1.38 x 104 1.10 x 104 7.77 x 103 2.45 x 104 0 hr 5

Cs-138 0 0 0 0 0 0 2.71 x 10 0 hr Ba-136m 0 0 0 0 0 0 1.50 x 103 1 hr Ba-137m 2.07 x 104 1.85 x 104 1.65 x 104 1.31 x 104 1.04 x 104 7.40 x 103 2.32 x 104 1 hr Ba-140 0 0 0 0 0 0 2.50 x 105 0 hr La-140 0 0 0 0 0 0 2.58 x 105 0 hr Ce-141 3.16 x 10-12 4.23 x 10-29 0 0 0 0 2.36 x 105 0 hr Ce-144 2.37 x 103 2.78 x 101 3.27 x 10-1 4.50 x 10-5 6.20 x 10-9 1.00 x 10-14 2.02 x 105 0 hr Pr-143 0 0 0 0 0 0 2.15 x 105 0 hr Pr-144m 3.69 x 101 4.57 x 10-1 5.36 x 10-3 8.11 x 10-7 1.12 x 10-10 1.99 x 10-16 3.02 x 103 1 day Pr-144 2.46 x 103 3.05 x 101 3.58 x 10-1 5.40 x 10-5 7.44 x 10-9 1.32 x 10-14 2.02 x 105 1 day Nd-144 7.41 x 10-11 7.50 x 10 -11 7.50 x 10 -11 7.50 x 10 -11 7.50 x 10 -11 7.50 x 10 -11 7.50 x 10 -11 15 yr U-235m 7.24 x 10-1 7.24 x 10-1 7.24 x 10-1 7.24 x 10-1 7.24 x 10-1 7.24 x 10-1 3.28 x 100 1 hr U-235 3.56 x 10-9 7.13 x 10-9 1.07 x 10-8 1.78 x 10-8 2.50 x 10-8 3.56 x 10-8 3.56 x 10-8 50 yr Np-239 0 0 0 0 0 0 2.72 x 106 0 hr Pu-239 7.24 x 10-1 7.24 x 10-1 7.24 x 10-1 7.24 x 10-1 7.24 x 10-1 7.24 x 10-1 7.25 x 10-1 1 yr Total 9.29 x 104 6.91 x 104 6.04 x 104 4.76 x 104 3.76 x 104 2.65 x 104 8.53 x 106 Notes:

Ci = Curies 2.4.13-14 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Table 2.4.13-2 (Sheet 1 of 2)

Radionuclide Release Concentrations (C0)

Radioactive Half-Life Initial Activity Initial Concentration Radionuclide (days) (Ci) (Ci/cm3)

H-3 4.51 x 103 1.25 x 102 3.30 x 100 C-14 2.09 x 106 1.37 x 10-1 3.62 x 10-3 Na-24 6.25 x 10-1 7.68 x 101 2.03 x 100 P-32 1.43 x 101 1.95 x 101 5.15 x 10-1 Cr-51 2.77 x 101 8.04 x 103 2.12 x 102 Mn-54 3.13 x 102 7.49 x 102 1.98 x 101 Mn-56 1.07 x 10-1 2.26 x 104 5.97 x 102 Fe-55 9.86 x 102 2.99 x 103 7.90 x 101 Fe-59 4.45 x 101 1.93 x 102 5.10 x 100 Co-58 7.08 x 101 1.21 x 103 3.20 x 101 Co-60 1.93 x 103 2.59 x 102 6.84 x 100 Ni-63 3.51 x 104 9.63 x 101 2.54 x 100 Cu-64 5.29 x 10-1 5.25 x 10-1 1.39 x 10-2 Zn-65 2.44 x 102 8.50 x 10-5 2.25 x 10-6 Rb-89 1.06 x 10-2 1.30 x 105 3.43 x 103 Sr-89 5.05 x 101 1.34 x 105 3.54 x 103 Sr-90 1.06 x 104 1.87 x 104 4.94 x 102 Sr-91 3.96 x 10-1 1.71 x 105 4.52 x 103 Sr-92 1.11 x 10-1 1.84 x 105 4.86 x 103 Y-90 2.67 x 100 1.94 x 104 5.13 x 102 Y-91m 3.45 x 10-2 5.30 x 104 1.40 x 103 Y-91 5.85 x 101 1.76 x 105 4.65 x 103 Y-92 1.48 x 10-1 1.86 x 105 4.91 x 103 Y-93 4.21 x 10-1 2.10 x 105 5.55 x 103 Zr-93 5.59 x 108 1.61 x 10-4 4.26 x 10-6 Zr-95 6.40 x 101 5.00 x 103 1.32 x 102 Nb-93m 4.97 x 103 1.41 x 10-4 3.73 x 10-6 Nb-95m 3.61 x 100 3.43 x 101 9.06 x 10-1 Nb-95 3.52 x 101 5.00 x 103 1.32 x 102 Mo-99 2.75 x 100 2.65 x 105 7.00 x 103 Tc-99m 2.51 x 10-1 2.35 x 105 6.21 x 103 Tc-99 7.78 x 107 1.01 x 10-2 2.67 x 10-4 Ru-103 3.93 x 101 2.24 x 105 5.92 x 103 Ru-106 3.68 x 102 8.63 x 104 2.28 x 103 Rh-103m 3.90 x 10-2 2.24 x 105 5.92 x 103 Rh-106 3.45 x 10-4 9.16 x 104 2.42 x 103 Ag-110m 2.50 x 102 4.31 x 102 1.14 x 101 Ag-110 2.85 x 10-4 6.46 x 100 1.71 x 10-1 2.4.13-15 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Table 2.4.13-2 (Sheet 2 of 2)

Radionuclide Release Concentrations (C0)

Radioactive Half-Life Initial Activity Initial Concentration Radionuclide (days) (Ci) (Ci/cm3)

Sb-124 6.02 x 101 1.18 x 102 3.12 x 100 Te-129m 3.36 x 101 7.21 x 103 1.90 x 102 Te-129 4.83 x 10-2 4.46 x 103 1.18 x 102 Te-131m 1.25 x 100 2.73 x 104 7.21 x 102 Te-131 1.74 x 10-2 4.80 x 103 1.27 x 102 Te-132 3.26 x 100 2.02 x 105 5.34 x 103 I-129 5.73 x 109 5.92 x 10-3 1.56 x 10-4 I-130 5.15 x 10-1 2.42 x 103 6.39 x 101 I-131 8.04 x 100 1.42 x 105 3.75 x 103 I-132 9.58 x 10-2 2.07 x 105 5.47 x 103 I-133 8.67 x 10-1 2.92 x 105 7.71 x 103 I-134 3.65 x 10-2 3.28 x 105 8.67 x 103 I-135 2.75 x 10-1 2.78 x 105 7.34 x 103 Xe-131m 1.18 x 101 3.98 x 102 1.05 x 101 Xe-133m 2.19 x 100 1.55 x 103 4.10 x 101 Xe-133 5.24 x 100 3.11 x 104 8.22 x 102 Xe-135m 1.06 x 10-2 3.73 x 104 9.84 x 102 Xe-135 3.81 x 10-1 5.93 x 104 1.57 x 103 Cs-134 7.53 x 102 3.01 x 104 7.95 x 102 Cs-135 8.40 x 108 7.11 x 10-5 1.88 x 10-6 Cs-136 1.31 x 101 1.00 x 104 2.64 x 102 Cs-137 1.10 x 104 2.45 x 104 6.47 x 102 Cs-138 2.24 x 10-2 2.71 x 105 7.16 x 103 Ba-136m 3.56 x 10-6 1.50 x 103 3.96 x 101 Ba-137m 1.77 x 10-3 2.32 x 104 6.12 x 102 Ba-140 1.27 x 101 2.50 x 105 6.61 x 103 La-140 1.68 x 100 2.58 x 105 6.82 x 103 Ce-141 3.25 x 101 2.36 x 105 6.24 x 103 Ce-144 2.84 x 102 2.02 x 105 5.34 x 103 1 5 Pr-143 1.36 x 10 2.15 x 10 5.68 x 103 Pr-144m 5.00 x 10-3 3.02 x 103 7.99 x 101 Pr-144 1.20 x 10-2 2.02 x 105 5.32 x 103 Nd-144 8.36 x 1017 7.50 x 10-11 1.98 x 10-12

-2 0 U-235m 1.81 x 10 3.28 x 10 8.67 x 10-2 U-235 2.57 x 1011 3.56 x 10-8 9.41 x 10-10 Np-239 2.36 x 100 2.72 x 106 7.19 x 104 Pu-239 8.79 x 106 7.25 x 10-1 1.91 x 10-2 Notes:

Ci = Curies 2.4.13-16 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Table 2.4.13-3 Computation of Aquifer Bulk Density Parameter Basis Geologic Formation Existing Fill/Residual Granular Weathered Soil Backfill Rock Benbolt Rockdell Fleanor Eidson Blackford Newala Material/USCS Limestone/ Limestone/ Limestone/

Symbol ML, MH, CH SW Siltstone Siltstone Limestone Siltstone Limestone Siltstone Dolomite Total Unit Weight, (lb/ft3) Lab testing 120 135 140 168 168 168 168 168 175 (wet density)

Natural Water Content, Lab testing 30 - - 1 1 1 1 1 1 w (% weight)

Specific gravity Lab testing 2.75 2.7 - 2.7 2.69 2.7 2.69 2.68 2.8 Saturated Bulk Density, Derived 1.92 2.16 2.24 2.69 2.69 2.69 2.69 2.69 2.80 sat (g/cm3)

Grain density, Assumed 2.75 2.7 2.7 2.7 2.7 2.7 2.7 2.7 2.8 g (g/cm3)

Total Porosity Derived 0.47 0.32 0.27 0.01 0.01 0.01 0.01 0.01 0.00 Dry Bulk Density, Derived 1.4 1.8 2.0 2.7 2.7 2.7 2.7 2.7 2.8 dry (g/cm3)

Notes:

Shaded cell highlights the value used in the analysis.

Relationships between porosity and density (Reference 2.4.13-5):

dry = (1 - n) g n = 1 - dry / g n = (g - sat) / (g - w) where n = total porosity, g = grain density, w = water density, dry = dry bulk density, and sat = saturated bulk density.

Total unit weight, , is assumed to be fully saturated.

USCS = Unified Soil Classification System ML = Silt, low plasticity MH = Silt, high plasticity CH = Clay, high plasticity SW = Sand, well-graded (diversified particle sizes) 2.4.13-17 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Table 2.4.13-4 (Sheet 1 of 2)

Distribution Coefficient (Kd) Test Results and Summary Statistics Bechtel Sample ID Cr Mn Fe Co Ni Zn Sr Zr Ru Ag Te Cs Ce Pu (OW/MP) mL/g mL/g mL/g mL/g mL/g mL/g mL/g mL/g mL/g mL/g mL/g mL/g mL/g mL/g 101-L/101 (145'-150') cm <1 1009 <1 420 115.1 564 2.8 <1 8.4 >4 >85 421 >267 >54 101-L/101 (145'-150') mm <1 324 <1 135 23.4 579 2.2 <1 0.01 <1 >22 353 >65 20.6 202-L/202 (156'-161') cm <1 4557 857 1207 79.4 323 <1 14.6 <1 76.5 1322 >19 33.7 202-L/202 (156'-156') mm <1 3720 835 322 86.1 341 <1 19 3.2 >119 2427 >17 39.1 202-L/202 (231'-236') cm <1 11161 538 247 1186 85.8 <1 7.2 <1 6 653 >15 30.9 202-L/202 (231'-236') mm <1 13795 1141 433 8671 134 <1 34.2 <1 57.5 1674 >13 32.8 409-U/409 (46'-51') mm <1 0.01 <1 132 22.7 70 7.9 <1 8.5 >806 >20 >4546 >71 19.5 409-L/409 (191'-196') cm <1 6043 <1 1187 230 564 29.5 <1 23.3 >872 9.1 560 42.2 >15 409-L/409 (191'-196') cm <1 11980 4.6 1878 326 615 29.4 <1 26.3 >865 5.3 594 >42 >15 409-L/409 (191'-196') mm <1 18820 8 2562 450 201 124 <1 30 >870 >32 1941 >39 4.7 415-U/415 (27'-35') cm <1 435 <1 341 59.3 207 30.1 4114 >10 >826 >21 >4318 >56 12.1 415-U/415 (27'-35') cm 1.2 296 <1 262 64.2 171 26 4788 5.7 >829 >21 >4252 >61 12.1 415-U/415 (27'-35') cm 1.4 353 <1 126 50.7 92.4 26.6 4533 >11 >823 >20 >4286 >68 14.7 415-U/415 (27'-35') mm 1.1 224 <1 249 67.2 809 26.7 3444 11.6 >774 >20 >4983 >69 14.3 415-U/415 (27'-35') mm 2.2 220 <1 298 75.6 423 26.7 3836 7.9 >967 >23 >5577 >87 >14 416-L/416 (66'-71') cm <1 3621 <1 444 49.4 66.9 3 75.3 2.7 >254 9.5 218 >45 >16 416-L/416 (66'-71') mm <1 1940 <1 902 69.3 316 4.8 111 4 >252 >19 2200 >45 >14 416-L/416 (111'-116') cm <1 690 <1 167 28.2 389 4.8 <1 237 >354 100.7 226 >75 >28 416-L/416 (111'-116') mm <1 360 <1 112 22.1 863 3.9 <1 2.1 >252 >20 939 >35 >14 418-U/418A (78'-85') cm <1 1141 <1 256 62.7 2725 5.1 <1 14 >857 4.4 89.1 >41 >12 418-U/418A (78'-85') cm <1 1225 <1 359 87 479 6.1 <1 10.6 >797 8 150 >41 >12 418-U/418A (78'-85') cm <1 969 <1 380 93.6 697 5.8 <1 13.7 >1051 10.8 146 >58 >15 418-U/418A (78'-85') mm <1 1351 <1 456 82 330 5.1 <1 10.9 >749 >23 688 >40 >11 418-U/418A (78'-85') mm <1 1321 <1 448 91.4 268 5.2 <1 8.5 >787 >25 678 >44 >13 419-U/419 (55'-62') cm <1 261 <1 355 19 153 1.7 <1 8.2 >813 6.7 99 >347 >21 419-U/419 (55'-62') mm <1 323 <1 800 49.9 907 1.7 <1 10.1 >825 27.9 1019 >294 >18 420-L/420 (132'-140') cm <1 1056 <1 40.6 20.7 333 <1 <1 5.8 782 >26 91.7 >267 >14 420-L/420 (132'-140') mm <1 566 <1 98.9 39.5 1505 1.3 <1 10.8 940 >27 585 >39 >14 423-U/423 (68'-76') cm <1 87 <1 152 79.2 726 15.5 <1 1.7 28 >23 1237 >35 >14 423-U/423 (68'-76') mm <1 40.9 <1 98 173 476 <1 <1 <1 29 >22 3099 >36 >13 2.4.13-18 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Table 2.4.13-4 (Sheet 2 of 2)

Distribution Coefficient (Kd) Test Results and Summary Statistics Bechtel Sample ID Cr Mn Fe Co Ni Zn Sr Zr Ru Ag Te Cs Ce Pu No. of test results 30 30 26 30 30 30 30 30 30 30 30 30 30 30 Minimum (mL/g) 0.5 0.5 0.5 40.6 19 66.9 0.5 0.5 0.5 0.5 4.4 89.1 13 4.7 Maximum (mL/g) 2.2 18820 8 2562 1207 8671 341 4788 237 1051 119 5577 347 54 Geometric mean (mL/g) 0.6 793.9 0.6 345.5 84.9 405.4 10.2 3.2 8.6 149.5 20.9 845.2 54.0 16.7 Median (mL/g) 0.5 989.0 0.5 357.0 72.5 449.5 6.0 0.5 10.1 784.5 21.5 813.5 44.5 14.5 Average (mL/g) 0.6 2929.6 0.9 534.3 155.3 818.4 42.7 697.1 18.6 547.0 29.7 1645.7 79.1 18.7 Notes:

Test results reported as <1 were assigned a value of 0.5 mL/g as a censored data. This is based on convention for censored data to be half of the detection limit (Reference 2.4.13-16).

Empty cells represent no data.

Test results reported as greater than a value (e.g., >20) were assigned the bounding value (e.g., 20) for statistical analysis.

Highlights indicate non-zero values used in the transport/dilution analysis.

2.4.13-19 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Table 2.4.13-5 (Sheet 1 of 3)

Transport/Dilution Analysis Parameters and Results Source Term Characteristics Dilution - No Sorption Sorption Parameters Dilution - With Sorption Minimum Minimum Decay Initial Initial Minimum Dilution River Minimum Dilution River Half-life(a) Constant(b) Activity(c) Concentration(d) ECL(e) Dilution Time Concentration(g) Kd(i) Dilution Time Concentration(g)

Radionuclide (days) (days-1) (Ci) (Ci/cm3) (Ci/cm3) Factor(f) (years) (Ci/cm3) C/ECL(h) (cm3/g) R(j) Factor(f) (years) (Ci/cm3) C/ECL(h)

H-3 4.51 x 103 1.54 x 10-4 1.25 x 102 3.30 x 100 1.00 x 10-3 6.02 x 106 0.94 5.49 x 10-7 5.49 x 10-4 0 1.0 6.02 x 106 0.94 5.49 x 10-7 5.49 x 10-4 C-14 2.09 x 106 3.32 x 10-7 1.37 x 10-1 3.62 x 10-3 3.00 x 10-5 5.71 x 106 0.94 6.34 x 10-10 2.11 x 10-5 0 1.0 5.71 x 106 0.94 6.34 x 10-10 2.11 x 10-5 Na-24 6.25 x 10-1 1.11 x 100 7.68 x 101 2.03 x 100 5.00 x 10-5 1.41 x 1057 0.17 1.43 x 10-57 0 0 1.0 1.41 x 1057 0.17 1.43 x 10-57 0 P-32 1.43 x 101 4.85 x 10-2 1.95 x 101 5.15 x 10-1 9.00 x 10-6 3.09 x 1012 0.62 1.67 x 10-13 0 0 1.0 3.09 x 1012 0.62 1.67 x 10-13 0 Cr-51 2.77 x 101 2.50 x 10-2 8.04 x 103 2.12 x 102 5.00 x 10-4 1.03 x 1010 0.72 2.07 x 10-8 4.14 x 10-5 0.6 19.0 1.04 x 1038 4.8 2.05 x 10-36 0 Mn-54 3.13 x 102 2.21 x 10-3 7.49 x 102 1.98 x 101 3.00 x 10-5 1.21 x 107 0.92 1.63 x 10-6 5.45 x 10-2 793.9 23801.0 * *

  • 0 Mn-56 1.07 x 10-1 6.48 x 100 2.26 x 104 5.97 x 102 7.00 x 10-5 5.77 x 10142 0.07 1.03 x 10-140 0 793.9 23801.0 * *
  • 0 Fe-55 9.86 x 102 7.03 x 10-4 2.99 x 103 7.90 x 101 1.00 x 10-4 7.27 x 106 0.93 1.09 x 10-5 1.09 x 10-1 0.6 19.0 7.51 x 109 15.3 1.05 x 10-8 1.05 x 10-4 Fe-59 4.45 x 101 1.56 x 10-2 1.93 x 102 5.10 x 100 1.00 x 10-5 7.57 x 108 0.79 6.74 x 10-9 6.74 x 10-4 0.6 19.0 1.13 x 1030 5.95 4.52 x 10-30 0 Co-58 7.08 x 101 9.79 x 10-3 1.21 x 103 3.20 x 101 2.00 x 10-5 1.36 x 108 0.84 2.35 x 10-7 1.17 x 10-2 345.5 10358.6 * * *
  • Co-60 1.93 x 103 3.59 x 10-4 2.59 x 102 6.84 x 100 3.00 x 10-6 6.46 x 106 0.94 1.06 x 10-6 3.53 x 10-1 345.5 10358.6 1.74 x 10111 973.32 3.93 x 10-111 0 Ni-63 3.51 x 104 1.97 x 10-5 9.63 x 101 2.54 x 100 1.00 x 10-4 5.75 x 106 0.94 4.43 x 10-7 4.43 x 10-3 84.9 2546.2 1.18 x 1016 1553.87 2.16 x 10-16 0 Cu-64 5.29 x 10-1 1.31 x 100 5.25 x 10-1 1.39 x 10-2 2.00 x 10-4 2.31 x 1062 0.16 6.01 x 10-65 0 0 1.0 2.31 x 1062 0.16 6.01 x 10-65 0 Zn-65 2.44 x 102 2.84 x 10-3 8.50 x 10-5 2.25 x 10-6 5.00 x 10-6 1.49 x 107 0.91 1.51 x 10-13 0 405.4 12154.3 * * *
  • Rb-89 1.06 x 10-2 6.54 x 101 1.30 x 105 3.43 x 103 9.00 x 10-4 * *
  • 0 0 1.0 * * *
  • Sr-89 5.05 x 101 1.37 x 10-2 1.34 x 105 3.54 x 103 8.00 x 10-6 4.42 x 108 0.8 8.01 x 10-6 1.00 x 100 10.2 306.8 6.32 x 10116 27.12 5.60 x 10-114 0 Sr-90 1.06 x 104 6.54 x 10-5 1.87 x 104 4.94 x 102 5.00 x 10-7 5.84 x 106 0.94 8.46 x 10-5 1.69 x 102 10.2 306.8 8.22 x 1011 231.98 6.01 x 10-10 1.20 x 10-3 Sr-91 3.96 x 10-1 1.75 x 100 1.71 x 105 4.52 x 103 2.00 x 10-5 3.44 x 1072 0.14 1.31 x 10-69 0 10.2 306.8 * *
  • 0 Sr-92 1.11 x 10-1 6.25 x 100 1.84 x 105 4.86 x 103 4.00 x 10-5 1.88 x 10140 0.07 2.59 x 10-137 0 10.2 306.8 * *
  • 0 Y-90 2.67 x 100 2.60 x 10-1 1.94 x 104 5.13 x 102 7.00 x 10-6 8.45 x 1026 0.33 6.06 x 10-25 0 54 1619.8 * *
  • 0 Y-91m 3.45 x 10-2 2.01 x 101 5.30 x 104 1.40 x 103 2.00 x 10-3 5.69 x 10254 0.04 2.46 x 10-252 0 54 1619.8 * *
  • 0 Y-91 5.85 x 101 1.18 x 10-2 1.76 x 105 4.65 x 103 8.00 x 10-6 2.54 x 108 0.82 1.83 x 10-5 2.29 x 100 54 1619.8 1.14 x 10252 67.46 4.07 x 10-249 0 Y-92 1.48 x 10-1 4.68 x 100 1.86 x 105 4.91 x 103 4.00 x 10-5 8.09 x 10120 0.08 6.07 x 10-118 0 54 1619.8 * *
  • 0 Y-93 4.21 x 10-1 1.65 x 100 2.10 x 105 5.55 x 103 2.00 x 10-5 1.69 x 1070 0.14 3.28 x 10-67 0 54 1619.8 * *
  • 0 Zr-93 5.59 x 108 1.24 x 10-9 1.61 x 10-4 4.25 x 10-6 4.00 x 10-5 5.71 x 106 0.94 7.45 x 10-13 0 3.2 96.9 5.53 x 108 91.48 7.69 x 10-15 0 Zr-95 6.40 x 101 1.08 x 10-2 5.00 x 103 1.32 x 102 2.00 x 10-5 1.87 x 108 0.83 7.08 x 10-7 3.54 x 10-2 3.2 96.9 3.33 x 1057 16.88 3.96 x 10-56 0 Nb-95m 3.61 x 100 1.92 x 10-1 3.43 x 101 9.06 x 10-1 3.00 x 10-5 1.39 x 1023 0.38 6.53 x 10-24 0 0 1.0 1.39 x 1023 0.38 6.53 x 10-24 0 Nb-95 3.52 x 101 1.97 x 10-2 5.00 x 103 1.32 x 102 3.00 x 10-5 2.42 x 109 0.76 5.46 x 10-8 1.82 x 10-3 0 1.0 2.42 x 109 0.76 5.46 x 10-8 1.82 x 10-3 Mo-99 2.75 x 100 2.52 x 10-1 2.65 x 105 7.00 x 103 2.00 x 10-5 3.38 x 1026 0.34 2.07 x 10-23 0 0 1.0 3.38 x 1026 0.34 2.07 x 10-23 0 Tc-99m 2.51 x 10-1 2.76 x 100 2.35 x 105 6.21 x 103 1.00 x 10-3 7.89 x 1091 0.11 7.87 x 10-89 0 0 1.0 7.89 x 1091 0.11 7.87 x 10-89 0 Tc-99 7.78 x 107 8.91 x 10-9 1.01 x 10-2 2.67 x 10-4 6.00 x 10-5 5.71 x 106 0.94 4.67 x 10-11 0 0 1.0 5.71 x 106 0.94 4.67 x 10-11 0 Ru-103 3.93 x 101 1.76 x 10-2 2.24 x 105 5.92 x 103 3.00 x 10-5 1.36 x 109 0.77 4.35 x 10-6 1.45 x 10-1 8.6 258.8 3.91 x 10121 21.98 1.51 x 10-118 0 2.4.13-20 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Table 2.4.13-5 (Sheet 2 of 3)

Transport/Dilution Analysis Parameters and Results Source Term Characteristics Dilution - No Sorption Sorption Parameters Dilution - With Sorption Minimum Minimum Decay Initial Initial Minimum Dilution River Minimum Dilution River Half-life(a) Constant(b) Activity(c) Concentration(d) ECL(e) Dilution Time Concentration(g) Kd(i) Dilution Time Concentration(g)

Radionuclide (days) (days-1) (Ci) (Ci/cm3) (Ci/cm3) Factor(f) (years) (Ci/cm3) C/ECL(h) (cm3/g) R(j) Factor(f) (years) (Ci/cm3) C/ECL(h)

Ru-106 3.68 x 102 1.88 x 10-3 8.63 x 104 2.28 x 103 3.00 x 10-6 1.08 x 107 0.92 2.10 x 10-4 7.01 x 101 8.6 258.8 4.41 x 1039 64.69 5.17 x 10-37 0 Rh-103m 3.90 x 10-2 1.78 x 101 2.24 x 105 5.92 x 103 6.00 x 10-3 1.11 x 10240 0.04 5.33 x 10-237 0 0 1.0 1.11 x 10240 0.04 5.33 x 10-237 0 Rh-106 3.45 x 10-4 2.01 x 103 9.16 x 104 2.42 x 103 N/A * *

  • 0 0 1.0 * *
  • N/A Ag-110m 2.50 x 102 2.77 x 10-3 4.31 x 102 1.14 x 101 6.00 x 10-6 1.46 x 107 0.91 7.81 x 10-7 1.30 x 10-1 149.5 4482.8 1.01 x 10203 231.75 1.13 x 10-202 0 Ag-110 2.85 x 10-4 2.43 x 103 6.46 x 100 1.71 x 10-1 N/A * *
  • 0 149.5 4482.8 * *
  • N/A Sb-124 6.02 x 101 1.15 x 10-2 1.18 x 102 3.12 x 100 7.00 x 10-6 2.29 x 108 0.82 1.36 x 10-8 1.94 x 10-3 0 1.0 2.29 x 108 0.82 1.36 x 10-8 1.94 x 10-3 Te-129m 3.36 x 101 2.06 x 10-2 7.21 x 103 1.90 x 102 7.00 x 10-6 3.13 x 109 0.75 6.08 x 10-8 8.69 x 10-3 20.9 627.6 2.23 x 10206 31.79 8.55 x 10-205 0 Te-129 4.83 x 10-2 1.44 x 101 4.46 x 103 1.18 x 102 4.00 x 10-4 5.13 x 10214 0.05 2.29 x 10-213 0 20.9 627.6 * *
  • 0 Te-131m 1.25 x 100 5.55 x 10-1 2.73 x 104 7.21 x 102 8.00 x 10-6 6.57 x 1039 0.24 1.10 x 10-37 0 20.9 627.6 * *
  • 0 Te-131 1.74 x 10-2 3.98 x 101 4.80 x 103 1.27 x 102 8.00 x 10-5 * *
  • 0 20.9 627.6 * *
  • 0 Te-132 3.26 x 100 2.13 x 10-1 2.02 x 105 5.34 x 103 9.00 x 10-6 2.23 x 1024 0.36 2.39 x 10-21 0 20.9 627.6 * *
  • 0 I-129 5.73 x 109 1.21 x 10-10 5.92 x 10-3 1.56 x 10-4 2.00 x 10-7 5.71 x 106 0.94 2.74 x 10-11 1.37 x 10-4 0 1.0 5.71 x 106 0.94 2.74 x 10-11 1.37 x 10-4 I-130 5.15 x 10-1 1.35 x 100 2.42 x 103 6.39 x 101 2.00 x 10-5 1.80 x 1063 0.15 3.56 x 10-62 0 0 1.0 1.80 x 1063 0.15 3.56 x 10-62 0 I-131 8.04 x 100 8.62 x 10-2 1.42 x 105 3.75 x 103 1.00 x 10-6 6.99 x 1015 0.51 5.37 x 10-13 0 0 1.0 6.99 x 1015 0.51 5.37 x 10-13 0 I-132 9.58 x 10-2 7.24 x 100 2.07 x 105 5.47 x 103 1.00 x 10-4 1.48 x 10151 0.07 3.69 x 10-148 0 0 1.0 1.48 x 10151 0.07 3.69 x 10-148 0 I-133 8.67 x 10-1 7.99 x 10-1 2.92 x 105 7.71 x 103 7.00 x 10-6 1.50 x 1048 0.2 5.14 x 10-45 0 0 1.0 1.50 x 1048 0.20 5.14 x 10-45 0 I-134 3.65 x 10-2 1.90 x 101 3.28 x 105 8.67 x 103 4.00 x 10-4 5.89 x 10247 0.04 1.47 x 10-244 0 0 1.0 5.89 x 10247 0.04 1.47 x 10-244 0 I-135 2.75 x 10-1 2.52 x 100 2.78 x 105 7.34 x 103 3.00 x 10-5 4.92 x 1087 0.11 1.49 x 10-84 0 0 1.0 4.92 x 1087 0.11 1.49 x 10-84 0 Xe-131m 1.18 x 101 5.85 x 10-2 3.98 x 102 1.05 x 101 1.00 x 10-8 2.80 x 1013 0.58 3.75 x 10-13 3.75 x 10-5 0 1.0 2.80 x 1013 0.58 3.75 x 10-13 3.75 x 10-5 Xe-133m 2.19 x 100 3.17 x 10-1 1.55 x 103 4.10 x 101 1.00 x 10-8 6.21 x 1029 0.3 6.59 x 10-29 0 0 1.0 6.21 x 1029 0.30 6.59 x 10-29 0 Xe-133 5.24 x 100 1.32 x 10-1 3.11 x 104 8.22 x 102 1.00 x 10-8 1.97 x 1019 0.44 4.16 x 10-17 0 0 1.0 1.97 x 1019 0.44 4.16 x 10-17 0 Xe-135m 1.06 x 10-2 6.53 x 101 3.73 x 104 9.85 x 102 N/A * *
  • N/A 0 1.0 * *
  • N/A Xe-135 3.81 x 10-1 1.82 x 100 5.93 x 104 1.57 x 103 1.00 x 10-8 1.05 x 1074 0.13 1.50 x 10-71 0 0 1.0 1.05 x 1074 0.13 1.50 x 10-71 0 Cs-134 7.53 x 102 9.21 x 10-4 3.01 x 104 7.95 x 102 9.00 x 10-7 7.82 x 106 0.93 1.02 x 10-4 1.13 x 102 845.2 25338.9 1.67 x 10279 957.62 4.75 x 10-277 0 Cs-135 8.40 x 108 8.25 x 10-10 7.11 x 10-5 1.88 x 10-6 1.00 x 10-5 5.71 x 106 0.94 3.29 x 10-13 0 845.2 25338.9 1.46 x 1011 23907.19 1.29 x 10-17 0 Cs-136 1.31 x 101 5.29 x 10-2 1.00 x 104 2.64 x 102 6.00 x 10-6 8.28 x 1012 0.6 3.19 x 10-11 5.32 x 10-6 845.2 25338.9 * *
  • 0 Cs-137 1.10 x 104 6.30 x 10-5 2.45 x 104 6.47 x 102 1.00 x 10-6 5.83 x 106 0.94 1.11 x 10-4 1.11 x 102 845.2 25338.9 3.30 x 1073 3602.79 1.96 x 10-71 0 Cs-138 2.24 x 10-2 3.09 x 101 2.71 x 105 7.16 x 103 4.00 x 10-4 * * *
  • 845.2 25338.9 * *
  • 0 Ba-136m 3.56 x 10-6 1.94 x 105 1.50 x 103 3.96 x 101 N/A * *
  • N/A 0 1.0 * *
  • N/A Ba-137m 1.77 x 10-3 3.91 x 102 2.32 x 104 6.13 x 102 N/A * *
  • N/A 0 1.0 * *
  • N/A Ba-140 1.27 x 101 5.46 x 10-2 2.50 x 105 6.61 x 103 8.00 x 10-6 1.19 x 1013 0.6 5.53 x 10-10 6.92 x 10-5 0 1.0 1.19 x 1013 0.60 5.53 x 10-10 6.92 x 10-5 La-140 1.68 x 100 4.13 x 10-1 2.58 x 105 6.82 x 103 9.00 x 10-6 1.41 x 1034 0.27 4.85 x 10-31 0 0 1.0 1.41 x 1034 0.27 4.85 x 10-31 0 Ce-141 3.25 x 101 2.13 x 10-2 2.36 x 105 6.24 x 103 3.00 x 10-5 3.79 x 109 0.75 1.64 x 10-6 5.48 x 10-2 54 1619.8 * * *
  • 2.4.13-21 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Table 2.4.13-5 (Sheet 3 of 3)

Transport/Dilution Analysis Parameters and Results Source Term Characteristics Dilution - No Sorption Sorption Parameters Dilution - With Sorption Minimum Minimum Decay Initial Initial Minimum Dilution River Minimum Dilution River Half-life(a) Constant(b) Activity(c) Concentration(d) ECL(e) Dilution Time Concentration(g) Kd(i) Dilution Time Concentration(g)

Radionuclide (days) (days-1) (Ci) (Ci/cm3) (Ci/cm3) Factor(f) (years) (Ci/cm3) C/ECL(h) (cm3/g) R(j) Factor(f) (years) (Ci/cm3) C/ECL(h)

Ce-144 2.84 x 102 2.44 x 10-3 2.02 x 105 5.34 x 103 3.00 x 10-6 1.31 x 107 0.91 4.09 x 10-4 1.36 x 102 54 1619.8 7.24 x 10113 147.71 7.37 x 10-111 0 Pr-143 1.36 x 101 5.097 x 10-2 2.15 x 105 5.68 x 103 2.00 x 10-5 5.39 x 1012 0.61 1.05 x 10-9 5.27 x 10-5 0 1.0 5.39 x 1012 0.61 1.05 x 10-9 5.27 x 10-5 Pr-144m 5.00 x 10-3 1.39 x 102 3.02 x 103 7.98 x 101 N/A * *

  • N/A 0 1.0 * *
  • N/A Pr-144 1.20 x 10-2 5.78 x 101 2.02 x 105 5.34 x 103 6.00 x 10-4 * * *
  • 0 1.0 * *
  • 0 Nd-144 8.36 x 1017 8.29 x 10-19 7.50 x 10-11 1.98 x 10-12 2.00 x 10-9 5.71 x 106 0.94 3.47 x 10-19 0 0 1.0 5.71 x 106 0.94 3.47 x 10-19 0 U-235m 1.81 x 10-2 3.84 x 101 3.28 x 100 8.67 x 10-2 N/A * *
  • N/A 0 1.0 * *
  • N/A U-235 2.57 x 1011 2.70 x 10-12 3.56 x 10-8 9.41 x 10-10 3.00 x 10-7 5.71 x 106 0.94 1.65 x 10-16 0 0 1.0 5.71 x 106 0.94 1.65 x 10-16 0 Np-239 2.36 x 100 2.94 x 10-1 2.72 x 106 7.19 x 104 2.00 x 10-5 4.74 x 1028 0.31 1.52 x 10-24 0 0 1.0 4.74 x 1028 0.31 1.52 x 10-24 0 Pu-239 8.79 x 106 7.89 x 10-8 7.25 x 10-1 1.92 x 10-2 2.00 x 10-8 5.71 x 106 0.94 3.35 x 10-9 1.68 x 10-1 16.7 501.6 2.90 x 109 473.18 6.60 x 10-12 3.30 x 10-4 (a) Values from References 2.4.13-8, 2.4.13-9, and 2.4.13-10 highlighted in yellow, green, and blue, respectively.

(b) Calculated as ln(2)/half-life.

(c) Initial activity is the peak activity value from Table 2.4.13-1.

(d) Calculated as initial activity divided by source term volume.

(e) Values from 10 CFR 20, Appendix B, Table 2, Column 2.

(f) Calculated using Equation 2.4.13-1 (Equation 4.41 of Reference 2.4.13-3).

(g) Calculated as Initial Concentration/Dilution Factor.

(h) Ratio of River Concentration to the effluent concentration limit (ECL). Values less than 10-6 are reported as zero.

(i) Kd = distribution coefficient; Based upon laboratory testing.

(j) R = retardation coefficient; Calculated using Equation 2.4.13-4.

Notes:

Ci = Curies N/A: Not applicable; no ECL available.

  • Indicates negligible concentrations in the Reservoir.

2.4.13-22 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Figure 2.4.13-1. Minimum Distance from the Power Block Area to the Clinch River Arm of the Watts Bar Reservoir 2.4.13-23 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Figure 2.4.13-2. Conceptual Model for Radionuclide Transport 2.4.13-24 Revision 1

Clinch River Nuclear Site Early Site Permit Application Part 2, Site Safety Analysis Report Source: Adapted from Reference 2.4.13-17 Figure 2.4.13-3. Alternate Conceptual Model for Radionuclide Transport 2.4.13-25 Revision 1

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