May 29, 2020

TECHNICAL MEMORANDUM CWMNW Arlington, OR

Analysis of Leachate Management Practices for the Chemical Waste Management of the Northwest Facility in Arlington, OR

Authors Arthur S. Rood, M.S., K-Spar, Inc Emily A. Caffrey, Ph.D., Radian Scientific, LLC Helen A. Grogan, Ph.D., Cascade Scientific, Inc. Colby D. Mangini, Ph.D., CHP, Paragon Scientific, LLC

Principal Investigator John E. Till, Ph.D., Risk Assessment Corporation

Submitted to Waste Management

RAC Technical Memorandum No. 5 - CWMNW Arlington - 2020

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1. Executive Summary

This technical memorandum provides a review of the radiation doses associated with the landfill leachate management methods and processes at the Chemical Waste Management of the Northwest’s (CWMNW) facility in Arlington, OR. Over time, precipitation that falls on the landfill surface and moisture that is entrained in the wastes moves by gravity through the waste mass – this is referred to as leachate. Leachate moves down by gravity through the waste mass, where it is intercepted by the primary liner and is then channeled by the leachate collection system to the leachate sumps where it is subsequently removed by dedicated pumps. The leachate is then pumped from the sumps via a hose and applied over the surface of three of the four landfill cells as a dust control measure. Over time, and depending on ambient weather conditions, the water phase in the leachate evaporates. CWMNW is located in an arid climate such that the application of leachate as dust control is the primary leachate management method. On days when leachate application to the landfill surface is not possible due to precipitation or snow cover, the leachate is pumped into a tanker truck that transports it to the on-site wastewater treatment plant (WWT-1) where it is offloaded into a storage tank for further treatment. Chemical flocculants are added to the leachate so that flocked solids precipitate to the base of the tank. The remaining liquid is passed through a carbon filter bed that removes most of the remaining contaminants and is then stored in a separate tank. Once the treated liquid passes confirmatory testing the liquids are pumped into one of two lined ponds. Periodically, the flocked solids and carbon filter media from WWT-1 are removed and disposed in the landfill. Leachate samples were taken from each of the four primary sumps on March 25-27, 2020. All samples were analyzed for naturally occurring radionuclides of the U-238 and Th-232 decay series, U-235, K-40, and tritium (H-3). These sampling results were used along with the volume of water and mass of disposed solids as the basis to determine the radionuclide concentrations in the wastes disposed in the landfill. The three exposure scenarios are considered in this review include: 1. Leachate applied as dust control on the surface of the landfill; 2. Disposal of the flocked solids and carbon filter media in the landfill; 3. Disposal of treated leachate in the evaporation ponds.

Receptors reviewed include landfill workers that operate on the landfill in full personal protective equipment, a laboratory worker, and the nearest current resident, located approximately 10,700 feet (more than two miles) from the edge of the landfill. This review finds that radiological doses from the leachate management practices at CWMNW are extremely low and do not suggest that any changes are necessary to the current leachate management methods. The maximum annual effective dose to a landfill worker who was assumed to spend 30 minutes per day on the landfill surface for 250 days per year from these practices was 0.22 mrem. Annual effective doses to the nearest resident were less than 0.005 mrem. The dose to a landfill worker from the disposal of flocked solids and carbon filter media from WWT-1 was also extremely low at 0.001 mrem. For comparison purposes, the average annual radiation dose from natural sources alone in the United States for an individual is approximately 310 mrem per year (NCRP 2009). This disposal scenario for the leachate treatment wastes is extremely unlikely in that it assumes all leachate from the landfill is treated through WWT-1; however, only a small fraction of the leachate is actually treated by the system because most of it is applied to the landfill surface as dust

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control. Further, the doses calculated for leachate being applied as dust control and the doses calculated for flocked solids and carbon filter media disposal are not additive as each scenario is evaluated independently assuming all of the radionuclides in the leachate are processed via that scenario. The highest calculated effective dose attributed to the landfill worker (0.22 mrem yr-1) is orders of magnitude less than the 25 mrem yr–1 recommended dose limit by the American National Standards Institute (ANSI 2009) for unrestricted release of soils from land containing Technologically Enhanced Naturally Occurring Radioactive Materials (TENORM), and the 100 mrem yr-1 public dose limit set by the Nuclear Regulatory Commission in 10 CFR § 20.1301.

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1. Introduction

This technical memorandum provides a review of the doses associated with the landfill leachate management methods and processes at the Chemical Waste Management of the Northwest’s (CWMNW) facility in Arlington, OR. This review of the doses associated with the leachate management practices has been completed using the March 2020 radioanalytical results of landfill leachate sampling conducted at the request of Oregon Department of Energy. Potential exposure pathways for radionuclides in the leachate are identified and potential doses to landfill employees and the public are evaluated. The CWMNW landfill is located about 11 km south of the Columbia River. Landfill Unit L- 14 is located is on the west side of the facility. Figure 1-1 shows the location of Landfill Unit L-14, the nearest CWMNW building with employees onsite (the CWMNW Laboratory), the nearest resident, and the meteorological station.

Figure 1-1. Location CWMNW, Landfill Unit L-14, the nearest CWMNW facility (CWMNW Laboratory), nearest resident, and the meteorological station.

Analysis of the Leachate Management Practices 3 at CWMNW

1.1. CWMNW Leachate Management System

Landfill Unit L-14 at CWMNW is a double lined Subtitle C design landfill with 4 lined leachate collection sumps that consist of a primary sump, secondary leak detection sump and tertiary leak detection sump collecting leachate from the current 86,490 m2 (21 acres) landfill. The landfill is divided into four cells with each cell designed to drain into a sump (Figure 1-2).

Figure 1-2. Landfill Unit L-14 showing the four cells (L-14 cell 1, L-14 cell 2, L-14 cell 3, and L- 14 cell 4), the sumps (S1, S2, S3, and S4), evaporations ponds (Pond-A and Pond-B), and nearby monitoring wells in the Selah formation.

Water infiltrates the landfill surface due to precipitation and to a lesser extent from using leachate for dust control by applying it to the top surface of the landfill. Those liquids filter that down through the entire waste mass and are conveyed by the primary liner to the leachate collection sumps at the base of the landfill. Leachate spraying is not expected to result in a large amount of water infiltration because spraying only occurs when evaporation is high. An alternate leachate management practice is utilized during periods of precipitation when leachate cannot be applied. An overview of the two leachate management practices is given in Figure 1-3.

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Leachate Spray System - Primary

L-14 cell 1 L-14 cell 2 L-14 cell 3 L-14 cell 4

Sump Sump Sump Sump

Wastewater Evaporation Storage Treatment Pond-A and Tank Plant Pond-B Water only sent to wastewater treatment Chemical flocculants plant during periods of added, particulates rain/snow and snow settle to base cover on landfill Solids/Filter media sent to landfill

Figure 1-3. Overview of leachate management at CWMNW.

When leachate is used for dust control, the leachate is pumped from the sumps via a hose and sprayed over the surface of the landfill where it evaporates. CWMNW is located in an arid climate which has 109 inches of dry pan evaporation per year. The spraying process continues until the area is adequately wetted. The approximate area of a single spray is illustrated in Figure 1-4. Once the area is adequately wetted, the sprayer is repositioned to a new location and the process repeated. Spraying is performed in areas of no disposal activity, mainly across cells L-14 cell 1, L-14 cell 2, and L-14 cell 3. Annually, the use of leachate for dust control will be distributed over all three cells.

Analysis of the Leachate Management Practices 5 at CWMNW

Figure 1-4. Google Earth view of the eastern portion of landfill L-14 showing the wetted spray area (dark grey regions) with the sprayer moving north to three other locations. Runoff from the spray operations is collected using the landfill internal stormwater collection system and sent to a separate lined stormwater pond at the north end of the current landfill. The calculated area of the southern wetted region is 337 m2.

An alternative leachate management practice is used when evapotranspiration is poor. On days where precipitation is occurring, or the ground is covered with snow, the leachate is pumped into a tanker truck that transports the leachate to the wastewater treatment plant-1 (WWT-1) where it is offloaded into a storage tank. Chemical flocculants are added to the leachate so that flocked solids precipitate to the base of the tank. The remaining liquid is passed through carbon filters and stored in a separate tank that is later pumped into one of two lined ponds (Pond-A, Pond-B) east of L-14 following compatibility and land disposal restriction (LDR) clearance testing. Periodically, the flocked solids and carbon filter media from the WWT-1 are removed and disposed in the landfill. This happens approximately six times per year.

2. Leachate Sampling and Analysis

Leachate samples were taken from each of the four primary sumps on March 25-27, 2020. All samples were analyzed for naturally occurring radionuclides of the U-238 and Th-232 decay series, U-235, K-40, and tritium (H-3). Tritium was measured using EPA Method 906.0 by Test America Inc, Denver CO. Uranium, thorium, radium, lead, and potassium isotopes were analyzed by ACZ R I S K A S S E S S M E N T C ORPORATION

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Laboratories Inc, Steamboat Springs CO. Uranium isotopes were analyzed using method EICHROM ACW03. The thorium isotopes were analyzed using method ESM 4506. Radium-226 was analyzed using method M903.1 and Ra-228 used method M904.0. Lead-210 was analyzed using method EICHROM OTW01, and K-40 was analyzed using gamma spectroscopy method EPA 901.1. A complete set of the analytical results is presented in Attachment 1.

2.1. Leachate Sampling Results

Analysis of the landfill leachate confirmed the presence of radionuclides of the uranium and thorium decay series, U-235, and K-40 (Table 1) in samples collected from each of the four disposal cell sumps (L-14 cells 1-4).

Table 1. Radionuclide Concentrations in Leachate Sump Water. Italicized Values Were Less than the Lower Limit of Detection (LLD) and Represent one-half the LLD value Radionuclide Concentration (pCi L–1) OAR 345- 050-0025 L-14 cell 1 L-14 cell 2 L-14 cell 3 L-14 cell 4 Average Table 1 Radionuclide Value U-238 358 6.53 73.9 6.5 111 1.0E+04 U-234 332 10.8 83 7.5 108 1.0E+04 Th-230 50.3 1.65 1.65 3.6 14.3 1.0E+02 Ra-226 1.2 0.85 0.75 0.35 0.79 1.0E+02 Pb-210 13.5 6.5 6.5 13 9.88 1.0E+02 Th-232 25.8 1.35 1.6 0.36 7.28 1.0E+02 Ra-228 24 10 10 10.5 13.6 5.0E+02 Th-228 32.9 0.672 0.573 8.04 10.5 1.0E+03 U-235 14.6 1.8 6.5 6.5 7.35 1.0E+04 K-40 126.4 427 474 474.72 375 5.0E+02 H-3 250 250 596 250 336.5 3.0E+07

Radionuclide concentrations in all leachate samples were significantly lower than OAR 345- 050-0025 Table 1 values. Uranium and thorium isotope concentrations were substantially higher in L-14 cell 1 compared to the other cells. Radium-226 concentrations were relatively constant, while only one Ra-228 concentration (in L-14 cell 1) was above the lower limit of detection (LLD). The K-40 concentration was lowest in L-14 cell 1 and relatively constant in the other cells. Tritium was also measured and was only above the reporting limit (500 pCi L–1) in the L-14 cell 3 sample. Tritium (12.33-year half-life) occurs naturally and is not a TENORM radionuclide. Its concentration was well below the drinking water standard of 20,000 pCi L–1 (40 CFR §141.66). The higher uranium and thorium concentrations relative to Ra-226 are reflective of the waste mass and age in cell L-14 cell 1 versus the new cells. Uranium in an oxidized state is soluble and also tends to have a lower soil-water partitioning coefficient compared to thorium and radium. Thorium is generally insoluble and has the highest soil-water partitioning coefficient of all the TENORM radionuclides. Sump water samples were not filtered and thus the increased concentrations of thorium isotopes in the L-14 cell 1 sump water may reflect thorium sorbed to suspended solids.

Analysis of the Leachate Management Practices 7 at CWMNW

Note that the Ra-228 concentration (half-life 5.75 years) is close to that of its parent (Th-232) in L- 14 cell 1 sample and probably reflects the ingrowth from Th-232 rather than some disposed source in L-14 cell 1.

3. Dose Assessment for Leachate Spraying

In this section the model for the release, transport, and dose assessment for spraying leachate as dust control on the top surface of landfill L-14 cells 1-3 is presented. The conceptual model is presented first, followed by the governing equations and model parameters.

3.1. Conceptual Model

The conceptual model (Figure 3-1) envisions the leachate volume collected in the sumps being sprayed over the area occupied by L-14 cells 1-3 during the period of a year. The radionuclides in the leachate used as dust control deposit on the landfill surface and are distributed in the first 1 cm of soil. A fraction of spray remains airborne and is subject to atmospheric transport and dispersion. A mass loading factor approach is used to compute the suspension of radionuclides in the surface layer to the air after the sprayer is moved. Over time, radionuclides move down through the soil column. In the surface layer, this movement is via physical movement of suspended solids and aqueous-phase radionuclides that have sorbed to soil particles. This process is termed percolation as described by Whicker and Kirchner (1987) in the PATHWAY model. Radionuclides that leave the surface layer are no longer susceptible to suspension or ingestion from soil depositing on the skin surface followed by hand- to-mouth ingestion. Radionuclides below the surface layer are subject to aqueous-phase leaching and transport. External exposure occurs from radionuclides in the surface and subsurface layers down to 15 cm. Radionuclides accumulate in the surface and subsurface layers until either the landfill is covered, or additional waste is disposed over the surface. For this assessment, it was assumed the landfill spraying continues for 50 years before a cover or additional waste is placed over landfill surface. Radionuclides that remain airborne during the spraying process can be inhaled both on and offsite. Potential receptors include a landfill worker who is present on the landfill surface, a worker in the CWMNW laboratory south of landfill L-14 (see Figure 1-1), and the nearest resident.

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Sprayer Fraction of spray to air (inhalation)

Inhalation from resuspension Soil ingestion

Leachate to soil (R) External External Exposure

Percolation (kp) Q1 1 cm

Q Aqueous-phase 2 Leaching (k ) 2 5 cm

Q3

Aqueous-phase

leaching (k3)

15 cm

Figure 3-1. Conceptual model for exposure assessment of leachate water application to the surface of landfill L-14.

3.2. Leachate Application Rate and Infiltration Rate

The flux of radionuclides entering the surface soil layer from leachate spray application is given by CV R = Li L (1) i A where –2 –1 Ri = leachate application rate for radionuclide i (pCi m yr ) –3 CLi = concentration in the leachate for radionuclide i (pCi m ) 3 –1 VL = annual volume of leachate (3,418.9 m yr ) A = area of L-14 cells 1-3 (58,185 m2)

To calculate the potential doses associated with spraying leachate on the landfill surface for dust control, the activity applied to the landfill surface in one year is required. The radionuclide concentrations measured in the leachate (Table 1) were used for this. Measured concentrations that were less than the LLD were set at one half the LLD for computation of the average. Aqueous-phase leaching in the subsequent layers is a function on the water infiltration rate. The water infiltration rate through the layers is given by

V I = L (2) A

Analysis of the Leachate Management Practices 9 at CWMNW where I = infiltration rate (m yr–1)

This value includes both water that infiltrates from precipitation and leachate applied to the landfill. Volumes of leachate collected in each of the four disposal cells (Table 2) vary with the area of the cell and volume of waste in place. The oldest cells (L-14 cells 1 and 2) have the most waste and backfill soil and no exposed liner. The newest cell (L-14 cell 4) has an exposed liner in the eastern portion of the cell and no spraying occurs on this cell. Water that falls as precipitation on L-14 cell 4 hits the exposed liner and runs down with minimal evaporation to the collection sump. Consequently, this cell collected the most leachate during 2019. A thin layer of waste over the liner (L-14 cell 3) will also allow more water to accumulate in the cell sump compared to a fully filled cell because as soon as water hits the impermeable liner it is channeled to the sump. Water that falls on the surface of older cells infiltrates into a thick layer of waste/soil and is held in the pore spaces where evaporation can remove a fraction of it. Note that the infiltration for L-14 cell 1 and 2 are about the same (~2.2 mm yr–1) while L-14 cell 3 is slightly greater (3.67 mm yr–1). These infiltration rates are comparable to those estimated at the Hanford reservation (DOE 2018) of 3.5 mm yr–1 and provide a good estimate of natural infiltration in the Arlington environment.

Table 2. Volume of Water Collected in Each of the L-14 Landfill Cells in 2019, Area of Each Cell, and Estimated Infiltration Volume Volume Area of cell Infiltration Cell (gal yr–1)a (m3 yr–1) (m2)b (m yr–1) L-14 cell 1 15,361 58.148 26,331.5 0.00221 L-14 cell 2 9,538 36.105 16,651.3 0.00217 L-14 cell 3 147,402 557.98 15,202.1 0.0367 L-14 cell 4 730,884 2,766.7 28,605.1 0.0967 Total 903,185 3,418.9 86,790 0.0394 Total, L-14 cell 1 through L-14 cell 3 ------58,185 0.0588 a. From J. Denson, email May 5, 2020. b. Calculated from the GIS coverages provided by Waste Management.

Dividing the total area of cells 1 through 3 (Table 2) by the spray area (337 m2, see Figure 1-4) indicates that about 172 individual spray locations would be needed to cover the entire land fill area. Assuming this takes place over a year, the sprayer is moved about every 2 days. Most of the liquid applied will evaporate before infiltrating because it is only performed during times when the pan evaporation rate is high. During periods of rain and snow, leachate water is sent to the WWT-1 and then released to either Pond A or Pond B. Thus, most of the water in the leachate collection system is from natural precipitation.

3.3. Governing Equations for Radionuclides Soil

The mass balance of radionuclides in each layer is adapted from the model in Whicker and Rood (2008) and described by the following series of first-order differential equations.

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dQij, =Ri −( k p + i) Q i, j i =1, j = 1 dt

dQij, =kp Q i, j− 1 −( k i , j + i) Q i , j i =1, j = 2 dt

dQij, =ki, j−− 1 Q i , j 1 −( k i , j + i) Q i , j i =1, j = 3 dt

dQij, =Ri −( k p + i) Q i, j + i−− 1 Q i 1, j i 1, j = 1 (3) dt

dQij, =kpij Q,− 1 −( k ij , + iij) Q , + iij − 1 Q − 1, i 1, j = 2 dt

dQij, = kQij,1− i, j− 1−(k i , j + i) Q i , j + i − 1 Q i − 1, j i 1, j = 3 dt where –2 –1 Ri = leachate application rate for radionuclide decay chain member i (atoms m yr ) –1 kp = percolation rate constant (0.6769 yr )

Qi,j = radionuclide inventory for radionuclide decay chain member i and layer j (atoms m–2) –1 ki,j = leach rate constant for radionuclide decay chain member i and layer j (yr )

Equations are written in terms of atoms because decay chains are involved. The conversion from activity to atoms is given by Q 0.037 dps/pCi Q = activity (4) atoms  where

Qactivity = radionuclide activity (pCi) –1  = decay rate constant (s )

Transport calculations were performed for an abbreviated decay chain because the short-lived members would not be present in the environment without the presence of their parent. The complete decay chain is given in the Section 3.7. The abbreviated U-238 decay chain was

U-238 → U-234 → Th-230 → Ra-226 → Pb-210.

The abbreviated Th-232 decay chain was

Th-232 → Ra-228 → Th-228.

No measurements of U-235 progeny were performed. This is reasonable because U-235 only represents a small fraction of the total uranium activity. Nevertheless, the short-lived progeny Th- 231 was included in the U-235 dose coefficient. Tritium was not included in the dose assessment because it will evaporate with the water and not accumulate in soil. The average concentration (336.5 pCi L–1) was lower than the drinking

Analysis of the Leachate Management Practices 11 at CWMNW water standard by a factor of 59, and drinking the leachate would bound any dose impacts from inhaling spray water. The percolation rate constant describes the movement of radionuclides from the surface soil to layer 2. The leach rate constant describes the movement of radionuclides from layers 2 to 3, and from layer 3 to deeper layers and is given by (Whicker and Rood 2008) I kij, = (5) Kdi b  Tj 1+  where –1 I = water infiltration rate (0.0588 m yr )  = moisture content (m3 m–3) 3 –1 Kdi = soil-water partitioning coefficient for radionuclide i (cm g )

Tj = thickness of layer j (m) 3 b = bulk density (g cm ).

The moisture content is a function of the pressure head and infiltration rate. It is calculated from a moisture characteristic curve using the fitting parameters described in van Genuchten (1980). The series of differential equations are embodied in the MCM model (Rood 2004; 2008) which was used to solve the system of equations and provide radionuclide inventories per square meter in each of the layers for the radionuclide application rate. The inventories in each layer were converted to soil concentrations by

Qij, Cij, = j (6) bkT k=1 –2 where Qi,j is the inventory has been converted to activity (pCi m ) for radionuclide i and layer j. Note that the concentration in layers 2 and 3 are really the average concentration from the surface to the depth of layers 2 and 3. This was done because external dose rate coefficients are in terms of the average concentration from the surface to the stated depth.

3.4. Atmospheric Transport

Radionuclides that remain suspended in the air from direct spraying are dispersed in the atmosphere. Dispersion in air was calculated using the U.S. Environmental Protection Agency model AERMOD v19191 (EPA 2015) and one-year of site-specific meteorological data (2010) obtained from the nearby meteorological tower (see Figure 1-1) operated by CWMNW. The meteorological data was processed with AERMET v12345 and the processed surface and upper air files were provided by CWMNW. For dispersion calculations, no deposition or plume depletion was assumed which maximizes the air concentration. An area source (254 m east-west, 149 m north-south) located in the center of the landfill was used to calculate annual dispersion factors. This area is appropriate and bounding because releases were not calculated for each individual spraying but collectively for a period of a year. The area (37,846 m2) is less than the area of cells L-14 cells 1-3 (58,185 m2) and thus confines releases to a smaller area resulting in higher air concentrations. The dispersion factor is the annual average airborne concentration (pCi m–3)

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divided by the source release rate (1 pCi s–1) and has units of s m–3. The product of the dispersion factor and the release rate gives the air concentration.

3.5. Model Parameters

Radionuclide-independent model parameters are summarized in Table 3 and element-specific soil-water partitioning coefficients are provided in Table 4. Area of application and calculated infiltration rates were discussed in the prior section. The thickness/depth of each layer was based on the thickness of external gamma dose coefficients. The surface layer assumes all the activity deposits in the 1-cm thick surface layer for external doses, but the removal rate constant does not depend on thickness. The percolation process discussed earlier can be applied to any layer of given thickness. Whicker and Kirchner (1987) assumed a 35-d half time (rate constant = ln(2)/35d × 365 d/1 yr = 7.2 yr–1) which means the surface compartment is depleted rapidly. For this application, a longer half-time of about a year was used based on U.S. Nuclear Regulatory Commission values, and thereby provides a bounding estimate of radionuclide residence times in the surface soil. The mass loading factor (MLF) is used to compute the concentration in air from the concentration in the surface soil. An upper bound value of 1.0×10–4 g m–3 from the RESRAD code (Yu et al., 2016) was adopted. This value is equivalent to 100 µg m–3 and all particulate matter is assumed to be 1 µm or less. The assumed particulate loading is greater than the 24-hour and annual standard for particulate matter less than 2.5 µm in air of 35 µg m–3 and 15 µg m–3, respectively Thus, the RESRAD value would certainly represent an upper bound value for this application. To determine an appropriate moisture content a material was selected that was reasonably conductive but also containing fine grained material like the Selah in which the base of the L-14 landfill is excavated into. The van Gentuchten fitting parameters (α and n) and properties (Ksat, r,

s) for sandy loam from Carsel and Parrish (1988) were selected to represent the backfilled soil and waste in the landfill. The calculated moisture content for an infiltration of 5.88 cm yr–1 was 0.139 m3 m–3.

Table 3. Model Parameters and Values Parameter Value Comments Area of application (m2) 86,790 Total area of landfill L-14 Infiltration (m yr–1) 0.0588 Average infiltration across L-14 landfill Bulk Density (g m–3) 1.76×106 Based on 2,970 lbs yd–3 as provided Geosyntec Consultants (2020) Layer 1 (surface) depth (m) 0.01 Corresponds to external exposure layer thickness Layer 2 depth (m) 0.05 Corresponds to external exposure layer thickness Layer 3 depth (m) 0.15 Corresponds to external exposure layer thickness Mass Loading Factor (g m–3) 1.0×10–4 RESRAD default Percolation rate constant for surface 0.6769 NRC (1975) as cited in Peterson (1983) layer (yr–1)

Analysis of the Leachate Management Practices 13 at CWMNW

Parameter Value Comments Moisture content (m3 m–3) 0.139 Calculated from infiltration rate and van Genuchten parameters for sandy loam Saturated hydraulic conductivity, Ksat 387.2 Carsel and Parrish (1988) (m yr–1) Saturated porosity, s 0.41 Carsel and Parrish (1988)

Residual moisture content, r 0.065 Carsel and Parrish (1988) alpha, α (m–1) 7.5 Carsel and Parrish (1988) n 1.89 Carsel and Parrish (1988) Dispersion factor, landfill worker (s m–3) 4.21E-5 Calculated with AERMOD Dispersion factor, CWMNW laboratory 9.91E-6 Calculated with AERMOD (s m–3) Dispersion factor, nearest resident 2.78E-7 Calculated with AERMOD (s m–3)

The soil-water partitioning coefficients (Kd) were obtained from several sources including the DOE-operated Hanford Reservation and Idaho National Laboratory (INL), default values in

RESRAD, and values from the literature. The Kd varies by orders of magnitude across the different media. For this reason, the geometric mean of the values presented in Table 4 were used in the MCM model.

Table 4. Linear Sorption Coefficients (Kd) and their Geometric Mean (GM) Sand Loam Clay RESRAD Hanford INL GM Element (mL g–1)a (mL g–1)a (mL g–1)a (mL g–1)b (mL g–1)c (mL g–1)d (mL g–1) U 35 15 1600 50 1 10 27 Th 3200 3300 5800 60000 1000 500 3500 Ra 500 36000 9100 70 14 500 657 Pb 270 16000 550 100 -- 270 577 K 15 55 75 20 -- -- 33 a. Sheppard and Thibault (1990) b. Yu et al. (2016) c. DOE (2018) d. Sondrup et al. (2018)

3.6. Dose Assessment

Annual effective dose was calculated to a landfill worker, a nearby worker at the laboratory south of landfill L-14, and the nearest resident. The landfill worker is exposed to surface and subsurface soil from inhalation, soil ingestion, and external exposure pathways. For this scenario, it is assumed that all the leachate is applied as dust control on the landfill surface, and no leachate is processed through the WWT-1 and subsequently discharged into the ponds. Inhalation and soil ingestion pathways for the landfill worker are considered unlikely because all personnel working inside the landfill footprint wear personnel protection equipment (PPE) that includes Tyvec suits, gloves, respirator, and safety glasses. Furthermore, worker safety and exposure protocols preclude anyone from being outside of the cab of a vehicle on the landfill R I S K A S S E S S M E N T C ORPORATION

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surface unless they are absolutely required to do so. For this assessment, no credit is taken for the respirator for the inhalation calculations and the worker is assumed to be outside the vehicle. The dose from inhalation to the landfill worker is given by

DIhi= C i,1  MLF  BR  EF  DCIh i (7) where

DIhi = inhalation dose for radionuclide i (mrem per year) –4 –3 MLF = mass loading factor (1×10 g m ) BR = breathing rate (m3 d–1) EF = exposure frequency (d yr–1) –1 DCIhi = inhalation dose coefficient for radionuclide i (mrem pCi ) –1 Ci,1 = concentration for radionuclide i in the surface layer (pCi g )

The soil ingestion pathway assumes some soil adheres to the gloves and Tyvec suit during operations and is inadvertently ingested during removal. The dose from soil ingestion to the landfill worker is given by

DIgi= C i,1  INGs  EF  DCIg i (8) where

DIgi = soil ingestion dose for radionuclide i (mrem per year) INGs = soil ingestion rate (g d–1) EF = exposure frequency (d yr–1) –1 DCIgi = inhalation dose coefficient for radionuclide i (mrem pCi ) –1 Ci,1 = concentration for radionuclide i in the surface layer (pCi g )

The soil ingestion rate was adjusted for the time actually spent on the landfill surface. The dose from external exposure to the landfill worker is given by

3 DExi= C i,, j  ET  DCEx i j (9) j=1 where

DExi = external exposure dose for radionuclide i (mrem per year) ET = exposure time (s yr–1) –1 –1 DCExi,j = external dose coefficient for radionuclide i and layer interval j (mrem-g pCi s ) –1 Ci,j = concentration for radionuclide i in layer interval j (pCi g )

The exposure scenario for the landfill worker (Table 5) assumed a single individual spent 0.5 hours per workday standing on top of a formerly applied area that had accumulated radionuclides for 50 weeks per year. Breathing and soil ingestion rates were taken from the TENORM assessment for the Blue Ridge Landfill (RAC 2019) and EPA (2016). No credit is taken for landfill worker personal protective equipment (PPE), which would substantially reduce dose from ingestion and inhalation pathways. For the laboratory worker and the nearest resident, only inhalation was considered because these individuals do not spend any time on the landfill surface and thus will not ingest landfill

Analysis of the Leachate Management Practices 15 at CWMNW surface soil or be exposed to external radiation. The air concentration these individuals were exposed to was calculated using the air concentration for the landfill worker and dispersion factors calculated with AERMOD. The laboratory worker was located 750 m upwind SSW from the center of landfill source. The nearest resident was 3.8 km SW from the center of the landfill source. The air concentration for these receptors was calculated as

DFk Ci, k=( C i ,1 MLF ) (10) DFLF where –3 Ci,k = annual average air concentration of radionuclide i at receptor k (pCi m ) –5 –3 DFLF = dispersion factor in the center of the landfill source (4.21×10 s m ) –6 –3 DFk = dispersion factor at receptor k (9.91×10 s m for laboratory worker and 2.78 × 10–7 s m–3 for the nearest resident)

Doses for these receptors are calculated using Equation 6, replacing Ci,1 × MLF with Ci,k and exposure parameters appropriate for the receptor. A final assessment of dose assumes the radionuclides emitted from the leachate used as dust control are dispersed in the atmosphere and inhaled by the landfill worker, laboratory worker, and nearest resident. The inhalation dose for this scenario is given by

CV DIn=Li L  ARF  RF  DF  BR  EF  DCIh (11) i, kCF k i where CF = 3.1536×107 s yr–1 ARF = airborne release fraction (0.001, unitless) RF = respirable (droplets that are in the respirable size fraction) fraction (0.87, unitless) i, k = radionuclide and receptor index.

The ARF represents the fraction of the radionuclides in the spray that remain airborne and can transport with the prevailing winds. The ARF and RF were obtained from Table 3-3 in DOE (1994) for venting of pressurized vessels under ambient temperatures. The highest ARF and RF values were used in the calculation.

Table 5. Exposure Scenario Parameters Parameter Value Reference and Comments Landfill Worker Hours per day 0.5 J. Densona Days per week 5 J. Densona Weeks per year 50 Assumed to have 2 weeks of vacation per year Days per year 250 Calculated Breathing rate (m3 hr–1) 1.8 RAC 2016 Breathing rate (m3 d–1) 0.9 Daily intake rate while spending and 0.5 hours per day on the landfill surface

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Parameter Value Reference and Comments Soil ingestion rate 20.6 Based on an 8-hr ingestion rate of 330 mg d–1 (EPA (mg d–1) 2016) adjusted for 0.5 hr d–1 spent on the landfill surface Hours per year 1000 Calculated Seconds per year 3,600,000 Calculated Laboratory Worker Breathing rate (m3 d–1) 7.4 Daily air intake while working. Based on 22.22 m3 d–1 in DOE (2011) multiplied by ratio of 8hrs/24hrs Hours per day 8 Assumed Days per week 5 Assumed Weeks per year 50 Assumed Days per year 250 Calculated Nearest Resident Breathing rate (m3 d–1) 22.2 DOE (2011) Days per year 365 Assumed a. Email from J. Denson to A.S. Rood May 15, 2020.

3.7. Dose Coefficients

Dose coefficients were obtained from Federal Guidance Report 15 (EPA 2019) for external exposure (Table 6) and U.S. Department of Energy Standard 1196 (DOE 2011) for inhalation and ingestion (Table 7). For external exposure, dose coefficients were those for an adult. For inhalation and ingestion, dose coefficients were for a reference individual as described in DOE (2011). The dose coefficients for a reference individual are slightly greater than those for an adult. Doses were calculated for an abbreviated decay chain because the short-lived progeny were assumed to be in secular equilibrium with their parent in the environment. This is done by summing the dose coefficients of the parent and progeny as indicated by the “+D” designation following the radionuclide. The inhalation and ingestion dose coefficients were taken from the tabulations in the RESRAD (Yu et al., 2016) code for DOE-Std-1196 dose coefficients. External dose coefficients from FGR-15 were converted from units of Sv-m3 (Bq-s)–1 to units of mrem-g (pCi-s)–1 using Equation 10 and the bulk density in Table 3.

Table 6. External Dose Coefficients from FGR 15 and Conversion from SI to Conventional Units 0-1 cm 0-5 cm 0-15 cm 0-1 cm 0-5 cm 0-15 cm Sv m3 Sv m3 Sv m3 mrem-g mrem-g mrem-g Radionuclide (Bq-s)–1 (Bq-s)–1 (Bq-s)–1 Fraction (pCi-s)–1 (pCi-s)–1 (pCi-s)–1 U-238 3.88E-22 7.06E-22 8.66E-22 2.53E-12 4.61E-12 5.66E-12 Th-234 4.92E-20 1.24E-19 1.56E-19 Pa-234m 4.98E-19 1.31E-18 2.05E-18 U-238+D 5.48E-19 1.43E-18 2.21E-18 3.58E-09 9.37E-09 1.44E-08 U-234 7.78E-22 1.57E-21 1.87E-21 5.08E-12 1.03E-11 1.22E-11 Th-230 2.06E-21 4.95E-21 6.15E-21 1.35E-11 3.23E-11 4.02E-11 Ra-226 4.01E-20 1.18E-19 1.67E-19

Analysis of the Leachate Management Practices 17 at CWMNW

0-1 cm 0-5 cm 0-15 cm 0-1 cm 0-5 cm 0-15 cm Sv m3 Sv m3 Sv m3 mrem-g mrem-g mrem-g Radionuclide (Bq-s)–1 (Bq-s)–1 (Bq-s)–1 Fraction (pCi-s)–1 (pCi-s)–1 (pCi-s)–1 Bi-214 8.89E-18 2.60E-17 4.19E-17 Pb-214 1.51E-18 4.38E-18 6.51E-18 Ra-226+D 1.04E-17 3.05E-17 4.86E-17 6.82E-08 1.99E-07 3.17E-07 Pb-210 7.28E-21 1.21E-20 1.25E-20 Bi-210 1.36E-19 3.74E-19 5.90E-19 Po-210 5.79E-23 1.68E-22 2.64E-22 Pb-210+D 1.43E-19 3.86E-19 6.03E-19 9.36E-10 2.52E-09 3.94E-09 Th-232 1.08E-21 2.34E-21 2.73E-21 7.06E-12 1.53E-11 1.78E-11 Ra-228 3.67E-22 5.28E-22 6.72E-22 Ac-228 5.20E-18 1.51E-17 2.39E-17 Ra-228+D 5.20E-18 1.51E-17 2.39E-17 3.40E-08 9.86E-08 1.56E-07 Th-228 1.09E-20 2.96E-20 3.92E-20 Ra-224 5.81E-20 1.73E-19 2.52E-19 Po-216 9.12E-23 2.65E-22 4.15E-22 Pb-212 8.11E-19 2.34E-18 3.34E-18 Bi-212 8.47E-19 2.38E-18 3.76E-18 Tl-208 1.89E-17 5.61E-17 9.32E-17 0.3594 Po-212 0.00E+00 0.00E+00 0.00E+00 0.6404 Th-228+D 8.52E-18 2.51E-17 4.09E-17 5.57E-08 1.64E-07 2.67E-07 U-235 8.90E-19 2.61E-18 3.69E-18 5.81E-09 1.71E-08 2.41E-08 Th-231 7.23E-20 1.84E-19 2.40E-19 U-235+D 9.62E-19 2.79E-18 3.93E-18 6.29E-09 1.83E-08 2.57E-08 K-40 1.12E-18 3.25E-18 5.25E-18 7.32E-09 2.12E-08 3.43E-08 a. See Equation 10 for conversion to conventional units in terms of soil mass.

DCEx (mrem-g [pCi-s]-1 ) = i (12) 3 -1 DCExib(Sv-m [Bq-s] ) 3700 mrem/pCi per Sv/Bq where 6 –3 b = bulk density (1.76×10 g m )

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Table 7. Inhalation and Ingestion Dose Coefficients from DOE-STD-1196 (DOE 2011) as Presented in RESRAD V2.7 (Yu et al., 2016) Inhalation Inhalation Solubility Ingestion Radionuclide (mrem pCi–1) Typea (mrem pCi–1) U-238+D 3.21E-02 S 2.13E-04 U-234 3.74E-02 S 2.15E-04 Th-230 3.85E-01 F 9.36E-04 Ra-226+D 3.82E-02 S 1.68E-03 Pb-210+D 4.01E-02 S 1.03E-02 Th-232 4.26E-01 F 1.03E-03 Ra-228+D 6.34E-02 S 5.92E-03 Th-228+D 1.75E-01 S 9.34E-04 U-235+D 3.38E-02 S 2.05E-04 K-40 3.28E-04 S 3.04E-05 a. RESRAD default solubility type of parent. In general, RESRAD default solubility type is the solubility type that has the highest dose coefficient. A less bounding assessment can be made using the solubility type recommended by ICRP (2001). Inhalation dose coefficients are based on a median particle size of 1 µm.

4. Dose Assessment for WWT-1 Pathways

This section presents the methods and modeling parameters used to estimate air concentrations and external dose rates from the disposal of flocked solids and carbon filter media from the on-site wastewater treatment plant WWT-1 at the Arlington Landfill. Environmental concentrations combined with exposure parameters and dose coefficients were used to estimate annual doses to landfill workers.

4.1. Disposal of Flocked Solids and Carbon Filter Media

Concentrations of radionuclides in the flocked solids and carbon filter media from the on-site wastewater treatment plant WWT-1 were determined using concentrations in the water (Table 1) and the volume of water (Table 2) to compute total activity (pCi) per radionuclide. The mass of solids disposed of in the landfill in 2019 was estimated by CWMNW to be 35 tons (3.18E+07 g). Radionuclides were assumed to be distributed homogeneously throughout this mass of waste to obtain a concentration (Table 8).

Table 8. Radionuclide Concentrations in Flocked Solids and Carbon Filter Media from the Wastewater Treatment Plant Radionuclide Concentration (pCi g-1) U-238 2.53E+00 U-234 2.73E+00 Th-230 4.37E-01 Ra-226 4.68E-02 Pb-210 1.28E+00 Th-232 1.08E-01

Analysis of the Leachate Management Practices 19 at CWMNW

Radionuclide Concentration (pCi g-1) Ra-228 1.15E+00 Th-228 7.72E-01 U-235 7.09E-01 K-40 5.04E+01

4.1.1. Exposure Scenario

The exposure scenario for disposal of flocked solids and carbon filter media from the on-site wastewater treatment plant assumes that there is one landfill worker that disposes of the waste twice per year. Complete pathways of exposure include inhalation of particulates, inadvertent soil ingestion, and external exposure. Exposure parameters are provided in Table 9.

Table 9. Exposure Parameters for Landfill Workers for Disposal of Flocked Solids and Carbon Filter Media Parameter Value Reference Breathing rate (m3 hr–1) 1.8 EPA 2011 Soil ingestion rate (mg d–1) 330 EPA 2016 Time per disposal (min) 7 CWMNW Number of disposals 6 CWMNW

4.1.2. Methods for Calculating Releases

This section describes the methods used to calculate releases to the atmosphere from the disposal of the flocked solids and carbon filter media.

4.1.2.1. Particulate Emissions and Inhalation and Ingestion Doses during Disposal Radionuclide emissions during disposal are based on the EPA emission model for aggregate handling and storage piles during drop loading operations as described in AP 42 Compilation of Air Pollutant Emission Factors (EPA 1995). Aggregate material is typically much drier and particulate aggregate is more easily dispersed in air than the flocked solids and carbon filter media that comprise the waste from the on-site wastewater treatment plant. Thus, modeling that material assuming it is aggregate results in the worst-case inhalation scenario and ensures doses are not underestimated. The exposure scenario is illustrated in Figure 4-1.

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Figure 4-1. Conceptual model of exposure for a landfill worker during disposal of the flocked solids and carbon filter media from the on-site wastewater treatment plant.

The emission factor is given by 1.3  U    2.2   (13) E = k(0.0016) 1.4  MC     2  where E = emission factor (kg released to air per Mg of material handled) U = wind speed (m s–1) MC = % moisture content k = particle size multiplier.

The product of the mass of radioactive material in each disposal of flocked solids and carbon filter media and the emission factor is the mass of radioactive material available for suspension in air. The amount of radionuclide release to the air is the product of the mass released to air and the representative radionuclide concentration. Thus,

1000 g Q = E  M C  (14) kg where Q = activity released to air (pCi) M = mass of one flocked solids and carbon filter media disposal (Mg) C = radionuclide concentration in flocked solids and carbon filter media (pCi g–1).

The total mass of flocked solids and carbon filter media disposed in 2019 was 35 tons (31.8 Mg), with an estimated bulk density of 2,970 lb yd-3 (1.76 g cm-3), yielding an estimated volume of

Analysis of the Leachate Management Practices 21 at CWMNW approximately 18 m3 per year, divided into six discrete disposals that are assumed to be placed into a single 25’ x 25’ disposal cell. The air concentration is then calculated by assuming the entire mass that is suspended is mixed in a mixing volume of air (defined below). The radionuclide concentration in air is then Q/V, where V is the volume of the mixing cell. The exposure scenario assumes the worker is exposed continuously until the material in air dissipates. The rate of removal from the mixing cell is described by the removal rate constant defined by

U K = (15) L where K = the removal rate constant (s–1) U = wind speed (m s–1) L = the length of the mixing cell that lies parallel to the direction of wind (m).

Assuming a square area source, the value of L is given by (A)1/2, where A is the surface area of the mixing cell. The change in concentration over time is described by the differential equation and solution dQ = -KQ dt (16) -Kt Q(t) = Qoe

where Qo is the initial activity in the mixing cell defined by Equation (13). The time-integrated air concentration that the worker is exposed to is calculated by

 QQQQ1  TIC=0 e−−kt dt = 0 − e kt = − 001 − = 0 (17)   0 ( ) V0  V K VK VK where TIC = the time-integrated concentration (pCi-s m–3) V = volume of the mixing cell (m3).

The area of the mixing cell was assumed to be the surface area of the disposal plus a buffer distance around the disposal that accounts for the distance to the worker in the cab of the heavy equipment. The surface area of the disposal is the disposal volume divided by the assumed average height of the pile. The mixing cell volume is the surface area (including buffer) × the difference between the height of the mixing cell and the average height of the pile.

V L = load +l Hload (18) 2 V = L (Hmc - Hload ) where 3 Vload = the volume of the load (~9 m per load, 2 loads per year)

Hload = height of the load after disposal (m)

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l = buffer distance (m)

Hmc = height of mixing cell (m).

The calculated waste pile height after dumping was approximately 0.2 m. The assumed height of the mixing cell was 2 m (6.56 ft), which allows suspended particles to be mixed with air on the side and on top of the pile. The buffer distance was assumed to be 3 m based on discussions with CWMNW. The inhalation dose from this exposure is given by n DINH= IR  TICjj  DCIh (19) j=1 where DINH = the inhalation effective dose for a disposal of flocked solids and carbon filter media (mrem) IR = inhalation rate (m3 s–1) –3 TICj = time-integrated concentration for radionuclide j (pCi-s m ) –1 DCIhj = inhalation effective dose coefficient for radionuclide j (mrem pCi ) n = number of radionuclides.

Ingestion effective doses during the disposal operation assume a given amount of the flocked solids and carbon filter media are ingested via adherence to skin and hand during a disposal, and later transferred to mouth. The nominal value for soil ingestion per day for a worker is adjusted for the worker’s exposure time during disposal of the waste, which was assumed to be 7 minutes. The ingestion effective dose is simply the product of the effective dose coefficient (in mrem pCi–1) and the amount of activity ingested. The amount of activity ingested is the soil ingestion rate adjusted for exposure time multiplied by the activity concentration of the waste (see section 4.1). The amount of material ingested is calculated by

330 mg 1 g 1 day 퐷 = × × × 0.12 hours × ∑푛 CA 퐷퐶퐼푔 (20) 푖푛푔 day 1000 mg 8 hours 푗=1 푗 푗 where

Ding = effective dose from ingestion (mrem) –1 CAj = weighted average concentration in TENORM for radionuclide j (pCi g ) –1 DCIgj = ingestion effective dose coefficient for radionuclide j (mrem pCi ).

Inhalation doses to off-site individuals are calculated using the amount of activity suspended into the air and a dispersion factor calculated using onsite meteorological data and the AERMOD model (EPA 2015). For the offsite resident, the 95th percentile one-hour dispersion factor for one year of data was used. Model parameters and calculated values are presented in Table 10. Inhalation and ingestion dose coefficients are discussed in section 3.7 and provided in Table 7.

Analysis of the Leachate Management Practices 23 at CWMNW

Table 10. Parameters for Emission Model during Disposal and Transport in Air Parameter Symbol Value Reference Average wind speed (m s-1) U 4.839 Mean value calculated from AERMOD surface file provided by CWMNW Moisture % MC 10 AP 42 (EPA 1995) Table 13.2.4- 1 in Section 13.2.4, mean value for clay in municipal landfills Wind speed multiplier K 0.48 AP-42 (EPA 1995) – assumes particles ≤15 µm are respirable

Volume of waste per Vload 3.0 Calculated disposal (m3)

-3 Bulk density (kg m ) b 1.76E+03 Geosyntec Consultants (2020)

Buffer distance (m) l 3.0 Assumed distance from edge of disposal pile to landfill worker in heavy equipment

Disposal pile height (m) Hload 0.31 Calculated waste thickness assuming all wastes are disposed in a single 25’ x 25’ cell

Mixing cell height (m) Hmc 2.0 Assumed height of air mixing cell Length of air mixing cell L 10.7 Calculated from Equation (6) (m) Air mixing volume (m3) V 22.2 Calculated from Equation (6) Distance to current nearest x 3,261 Distance to nearest current offsite resident (m) resident estimated from Google Earth imagery Removal rate constant (s-1) K 0.5 Calculated using Equation (3) Emission rate (kg released E 1.19E-03 Calculated using Equation 1 to air per load) from AP-42 (EPA 1995)

4.1.3. External Dose During Waste Disposal

Using the representative inventory in Table 8 and the exposure geometry depicted in Figure 4-1, dose factors for Ra-226 and Ra-228 were calculated using the MicroShield code (Grove Engineering, Inc. 2013). These values can be used to estimate dose to the landfill worker operating the heavy equipment during disposal. The general equation for estimating dose from external exposure is D = DR ET  DF (21) where DR = the dose rate on contact with the material ET = the exposure time (hours) DF = distance factor (unitless).

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Table 11. External Dose Calculation Parameters Parameter Value Units Ra-226 dose factor 5.65E-05 mrad hr-1 per pCi g-1 Ra-228 dose factor 8.32E-05 mrad hr-1 per pCi g-1 Exposure rate at waste handler position 9.80E-05 mrem hr-1 from unshielded wastea a. Assumes 1 mrad = 1 mrem. No credit taken for shielding provided by heavy equipment nor PPE worn by landfill worker.

4.1.4. Radon Exposure and Dose

Emission of radon from the Landfill is likely a continuous exposure situation, and therefore, the assessment focuses on emissions after waste is placed and covered in its final state of disposition. Radon-222 emissions from the landfill resulting from disposal of the flocked solids and carbon filter media were calculated using Nuclear Regulatory Commission models and methods for assessment of uranium mill tailings (Rogers et al. 1984). A diffusion model is used to first calculate radon flux from the surface of uncovered compacted waste containing the flocked solids and carbon filter media and other chemical and hazardous wastes. The flux from the bare surface is given by

   J =104 C E D tanh x  (22) t b t  t   Dt  where –2 –1 Jt = flux from the surface of waste layer in the disposal cell (pCi m s ) C = Ra-226 concentration in flocked solids and carbon filter media (pCi g–1) 2 –1 Dt = radon diffusion coefficient in flocked solids and carbon filter media (m s )  = radon decay constant (2.1×10–6 s–1) -3 b = bulk density of waste (g cm ) E = Rn-222 emanation coefficient

xt = thickness of waste (cm).

The Ra-226 concentration is the measured concentration in the leachate sump water. The flocked solids and carbon filter media are placed in the landfill and covered with other RCRA hazardous wastes, and ultimately an infiltration-reducing cover is installed. The radon flux after burying and covering the waste is given by

−bc xc 2Jt e J c = −2bc xc (1+ at ac tanh(bt xt ))+ (1− at ac tanh(bt xt ))e

bi =  Di ,i = c or t (23) 2 ai = Di (1− (1− k)mi )  1 1  −2   mi =10 MP −   b s  where

Analysis of the Leachate Management Practices 25 at CWMNW

–2 –1 Jc = radon flux from the disposal cell surface (pCi m s ) –3 s = particle density (g cm )  = porosity MP = dry weight percent moisture (g of water g-1 of dry soil × 100) k = 0.26 pCi cm–3 in water per pCi cm–3 in air

mi = moisture saturation fraction for waste (i=t) or cover (i=c).

The radon diffusion coefficient is given by

2 5 Di = 0.07 exp− 4(m − m − m ) (24)

The flux at the surface can be compared to the limit applied to uranium mill tailings disposal cells of 20 pCi m–2 s–1. Doses from radon are dependent on the radon progeny concentrations in air that exist in various levels of equilibrium with radon. Doses were estimated using the working level (WL) and a conversion of 760 mrem per working-level month (Yu et al. 2001). The WL is defined as any combination of short-lived radon progeny in one liter of air that will result in the emission of 1.3×105 MeV of potential alpha energy. One WL equals 100 pCi L–1 of radon in air with all short- lived progeny in equilibrium. The WL is related to the equilibrium equivalent concentration (EEC) and given by NCRP (1988)

EEC = 0.105A + 0.516B + 0.379C (25) where A, B, and C are the concentrations of Po-218, Pb-214, and Bi-214, respectively. For these calculations, we assume worst-case conditions where radon progeny are in equilibrium with radon. If A, B, and C are measured in pCi L–1, then 1 WL = EEC/100. Assuming progeny are in equilibrium with radon (a worst-case assumption) and 1 pCi L–1 radon concentration, then the EEC is 1 EEC per pCi L–1. The working level month (WLM) and dose from radon is given by

hours exposed WLM= WL 170 hours (26) mrem D=760 WLM WLM

Radon model parameters are presented in Table 12.

Table 12. Radon Model Parameters Parameter Symbol Value Notes Waste thickness (m) xt 0.31 Calculated waste thickness assuming all wastes are placed in single 25’ × 25’ disposal cell Cover thickness (m) xc 32.83 Provided by CWMNW

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Parameter Symbol Value Notes Dry weight % moisture, waste MP 5.19 Calculated assuming 1 mm of infiltration per year Dry weight % moisture, cover MP 5.19 Calculated assuming 1 mm of infiltration per year -3 Bulk density, waste (g cm ) b 1.76 Geosyntec Consultants (2020) -3 Bulk density, cover (g cm ) b 1.76 Geosyntec Consultants (2020) Porosity, waste  0.41 CWMNW Updated Hydrogeologic Conceptual Site Model Report (CH2MHILL 2008) Porosity, cover  0.41 CWMNW Updated Hydrogeologic Conceptual Site Model Report (CH2MHILL 2008)

Particle density, waste s 2.99 Calculated using s = b/(1-)

Particle density, cover s 2.99 Calculated using s = b/(1-) Radon emanation coefficient E 0.2 Typical value for uranium mill tailings (see text) Ra-226 concentration (pCi g-1) C 4.4E-04 Calculated based on the total Ra- 226 inventory placed in one disposal block Surface area of waste disposals (m2) A 58.06 Assumes all wastes are placed in single 25’ × 25’ disposal cell

Waste thickness and cover thickness were based on information provided by CWMNW. Radium-226 (the radon source) was assumed to be uniformly distributed throughout the waste. The worst-case Ra-226 concentration was estimated by placing the entire Ra-226 inventory (1.49E-06 Ci) in one 7.6 m × 7.6 m disposal cell and applying the bulk density of 1.76E+03 kg m–3. Radon flux generally increases with waste thickness until the radon diffusion time is sufficient to result in decay of radon generated in the lower levels before exiting the top. For the waste over the flocked solids and carbon filter media, the average waste thickness (~33 m) was used. Thicker covers will attenuate radon flux allowing for decay of radon within the cover before emission to the surface. The waste was assumed to be relatively dry with a dry weight percent moisture of ~5%. Rogers et al. (1984) showed that radon diffusion coefficients decrease with moisture saturation. A doubling of the moisture saturation results in a decrease in the radon diffusion coefficient by a factor of 2 or more (see Figure 12 in Rogers et al. [1984]). Typically, covers for uranium mill tailings have moisture contents ranging from 6% to 11%. Thus, a dry weight percent moisture of just over 5% is considered worst-case because it maximizes fluxes. The radon emanation coefficient was assumed similar to uranium mill tailings, and a value of 0.2 was selected based on Figure 15 in Rogers et al. (1984).

Analysis of the Leachate Management Practices 27 at CWMNW

4.2. Potential Releases from the Lined Evaporation Ponds

Leachate that is treated through the on-site WWT-1 is discharged to the lined evaporation ponds located east of landfill L-14. Release of radionuclides from the water through volatilization is impossible for all radionuclides except radon because none of the radionuclides are volatile. Moreover, the on-site WWT-1 will remove radionuclides in the suspended solids and dissolved- phase radionuclides in the carbon media. Thus, radionuclides entering the evaporation pond will much less than what is emitted through the leachate being applied to the landfill surface for dust control. Additionally, the on-site evaporation ponds only receive treated leachate during times of limited evaporation. Radon release from the pond is limited by the water layer that serves as a diffusion barrier. The bulk radon diffusion in air is 0.11 cm2 s–1 while the diffusion coefficient in water is four orders of magnitude lower (10–5 cm2 s–1) (Nielson and Rogers 1982). Thus, most radon will decay in the water before being released to air. Typically, the on-site evaporations ponds have liquid volumes in them throughout the year thereby eliminating the emission of the miniscule fraction of radionuclides that pass through the on-site WWT-1 system. To assess the unlikely case where a potential release of radionuclides contained in water droplets generated by wind blowing across the pond surface, Equation 3-13 in DOE (1994) has been used.

−11 0.098039u CCair= water (3.0 10) 10 (27) where –3 Cair = concentration in air (pCi m ) –3 Cwater = solute concentration in water (pCi m ) u = windspeed (m s–1)

Assuming the mean windspeed of 4.839 m s–1 measured at the meteorological station, the concentration in air is a factor of 8.9×10–11 times the concentration in the water. Using the average Th-230 concentration measured in leachate of 14,300 pCi m–3 (14.3 pCi L–1) gives an air concentration of 1.28×10–6 pCi m–3. This concentration estimate is used but is unlikely as this assumes the leachate is discharged untreated into the ponds. Concentrations therefore would be even lower if treated leachate is assumed. The air concentration standard for Th-230 in Table 3 of OAR 345-050-0035 is 0.08 pCi m–3 (assuming soluble Th which is the most limiting value). The OAR 345-050-0035 Table 3 is 62,500 times greater than the estimated concentration from the pond using the unlikely case. Thus, the pond water is not of concern for exposure or dose. For these reasons, dose impacts from leachate management practices at CWMNW are bounded by leachate being applied as dust control and disposal of the flocked solids and carbon filter media. Therefore, doses from the release of radionuclides from the evaporation ponds are negligible and not considered further.

5. Results

Effective doses for a landfill worker and offsite receptors are presented in this section for leachate spraying and disposal of the flocked solids and carbon filter material in the landfill. It is

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important to note that the doses calculated for leachate being applied as dust control and the doses calculated for flocked solids and carbon filter media disposal cannot be added because each scenario is evaluated independently assuming all of the radionuclides in the leachate are processed via that scenario.

5.1. Leachate Applied as Dust Control

The leachate applied as dust control scenario theoretically results in the build-up of radionuclides in soil over time. This report assumes this material is suspended into air and contributes to external exposure for a person standing on the landfill surface. As discussed earlier, this scenario is considered unlikely as all landfill workers wear respiratory protection while working on the landfill. The MCM model was used to compute the build-up in soil over a 50-year period. Concentrations in soil as a function of time are presented first followed by the dose estimates for the three scenarios.

5.1.1. Concentrations in Soil

The build-up of radionuclides in soil over a 50-year period are illustrated in Figure 5-1, Figure 5-2, and Figure 5-3 for the surface, 0-5 cm layer, and 0-15 cm layer respectively. The surface layer (Figure 5-1) shows radionuclide concentrations reach equilibrium in about 5 years whereas in the deeper layers (Figure 5-2 and Figure 5-3) radionuclides continue to accumulate. Concentrations in the surface layer at equilibrium are roughly equal to those in the 0-5 cm layer at 7 years whereas concentration in the third layer are about a factor of 3 less than the surface layer. Note that the Ra- 226 concentrations are substantially below 5 pCi g–1 (~0.004 pCi g–1 in layer 1 and 2, and 0.001 in layer 3). Furthermore, the Ra-226 standard in soil is 5 pCi g–1 in the first 15 cm of soil (40 CFR 192.12). Thus, the standard should be compared to the 0-15 cm layer (layer 3). The Ra-226 concentration in layer 3 is 5000 times less than Ra-226 standard.

Analysis of the Leachate Management Practices 29 at CWMNW

Figure 5-1. Radionuclide concentrations in soil layer 1 (0-1 cm) as a function of time.

Figure 5-2. Radionuclide concentrations in soil layer 2 (0-5 cm) as a function of time.

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Figure 5-3. Radionuclide concentrations in soil layer 3 (0-15 cm) as a function of time.

5.1.2. Dose Estimates

The total annual effective dose from soil receiving radionuclides from leachate used as dust control on the landfill surface after 50 years of buildup (Table 13) was 0.22 mrem for the landfill worker who spends time on the landfill. External exposure accounts for ~96% of the dose and K- 40 contributes about 62% of the total dose; however, K-40 is naturally occurring and generally does not present a health impact. Typical background concentrations of K-40 in soils range from 10 to 30 pCi g–1 (Huffert et al., 1994), which is greater than the concentrations seen in Figure 5-1 through Figure 5-3. The total annual effective dose as a function of buildup time (Figure 5-4) shows that after 15- years of buildup, doses are about half of what the dose is at 50-years. If at any time during the buildup period, the landfill is covered with a sufficient amount of cover material (waste/soil of about a meter thick) then doses would drop to zero and the buildup would start over again.

Table 13. Estimates of Annual Effective Dose to a Worker at Landfill L-14 from Soil Receiving Radionuclides During Leachate Spraying After a 50-year Buildup Period Inhalation Ingestion External Total Radionuclide (mrem) U-238+D 3.95E-04 6.01E-04 1.65E-02 1.75E-02 U-234 4.48E-04 5.90E-04 1.61E-05 1.05E-03 Th-230 6.09E-04 3.39E-04 1.02E-05 9.58E-04 Ra-226+D 3.36E-06 3.39E-05 4.30E-03 4.34E-03 Pb-210+D 4.20E-05 2.46E-03 3.26E-04 2.83E-03 Th-232 3.43E-04 1.90E-04 2.42E-06 5.35E-04

Analysis of the Leachate Management Practices 31 at CWMNW

Inhalation Ingestion External Total Radionuclide (mrem) Ra-228+D 8.87E-05 1.90E-03 1.96E-02 2.16E-02 Th-228+D 2.19E-04 2.67E-04 3.34E-02 3.39E-02 U-235+D 2.74E-05 3.81E-05 2.03E-03 2.10E-03 K-40 1.36E-05 2.89E-04 1.37E-01 1.37E-01 Total 2.19E-03 6.71E-03 2.13E-01 2.22E-01

Figure 5-4. Annual effective dose to the landfill worker as a function of buildup time assuming no cover or additional waste is placed on the surface.

Doses to the laboratory worker and nearest resident from soil exposed to leachate used as dust control were significantly lower, as external exposure is not a viable pathway and dispersion and dilution reduces concentrations in air at the receptor location. Total annual effective dose for the laboratory worker was 0.0042 mrem and 0.00047 mrem for the nearest resident. Thorium-230 was the primary contributor to total dose.

Table 14. Estimates of Annual Effective Dose to a Laboratory Worker and Nearest Resident from Soil Exposed to Leachate used as Dust Control Laboratory Worker Nearest Resident Radionuclide (mrem) U-238 7.6E-04 8.4E-05 U-234 8.6E-04 9.5E-05 Th-230 1.2E-03 1.3E-04

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Laboratory Worker Nearest Resident Radionuclide (mrem) Ra-226 6.5E-06 7.1E-07 Pb-210 8.1E-05 8.9E-06 Th-232 6.6E-04 7.3E-05 Ra-228 1.7E-04 1.9E-05 Th-228 4.2E-04 4.7E-05 U-235 5.3E-05 5.8E-06 K-40 2.6E-05 2.9E-06 Total 4.2E-03 4.7E-04

Dose from the fraction of radionuclides that become airborne during application (Table 15) were considerably lower than doses from soil on the landfill surface exposed to leachate used as dust control. Total annual doses were 1.4×10–4 mrem, 3.4×10–5 mrem, and 3.8×10–6 mrem for the landfill worker, laboratory worker, and nearest resident respectively. All doses are substantially less than the 25 mrem yr–1 dose limit recommended by the American National Standards Institute (ANSI 2009) for unrestricted release of TENORM contaminated land, and well below the 100 mrem per year public dose limit set by the Nuclear Regulatory Commission in 10 CFR §20.1301.

Table 15. Annual Inhalation Doses from Radionuclides in Leachate that Remain Suspended in Air During Application

Radionuclide Landfill Laboratory worker Nearest resident (mrem) U-238 2.6E-05 6.2E-06 6.8E-07 U-234 2.9E-05 7.0E-06 7.8E-07 Th-230 3.9E-05 9.5E-06 1.1E-06 Ra-226 2.2E-07 5.2E-08 5.8E-09 Pb-210 2.8E-06 6.9E-07 7.6E-08 Th-232 2.2E-05 5.4E-06 5.9E-07 Ra-228 6.2E-06 1.5E-06 1.7E-07 Th-228 1.3E-05 3.2E-06 3.5E-07 U-235 1.8E-06 4.3E-07 4.8E-08 K-40 8.8E-07 2.1E-07 2.4E-08 Total 1.4E-04 3.4E-05 3.8E-06

5.2. Doses from Disposal of WWT-1 Wastes

Effective doses to landfill workers and the public from the disposal of flocked solids and carbon filter media from WWT-1 system are presented in Table 16. The doses assume the same worker performs all disposals. The doses are very low. In the U.S. the average annual radiation dose to an individual from natural sources alone is approximately 310 mrem per year (NCRP 2009). Further, these effective doses are substantially less than the 25 mrem yr–1 dose limit recommended by the American National Standards Institute (ANSI 2009) for unrestricted release of soils from

Analysis of the Leachate Management Practices 33 at CWMNW land containing TENORM, and well below the 100 mrem per year public dose limit set by the Nuclear Regulatory Commission in 10 CFR §20.1301.

Table 16. Annual Effective Doses to the Landfill Worker and Nearest Current Resident During Disposal of Flocked Solids and Carbon Filter Media Dose to Landfill Worker Dose to Nearest Current Resident Pathway (mrem) Inhalation 1.4E-03 4.4E-10 Ingestion 2.2E-04 NA External 2.3E-05 NA Total 1.6E-03 4.4E-10

Radon concentration, working level month (WLM), and dose for the landfill worker, nearest current onsite resident, and potential future residents are presented in Table 17. The doses represent annual doses. The flux at the surface was 2.3×10–16 pCi m-2 s-1, substantially below the 20 pCi m-2 s-1 limit.

Table 17. Radon Concentration, WLM, and Dose Radon Concentration Pathway WLM Dose (mrem) (pCi L-1) Landfill worker 1.63E-16 1.96E-17 1.5E-14 Current offsite resident 3.78E-18 1.50E-18 1.1E-15 Future offsite resident 3.78E-18 1.87E-18 1.4E-15 Future onsite resident 1.63E-16 8.05E-17 6.1E-14

6. Limiting Leachate Concentrations

As the landfill worker is the limiting exposure scenario, the dose received by the landfill worker is used to determine the maximum radionuclide concentration in leachate such that the annual public dose limit of 100 mrem is not exceeded. Table 18 shows the maximum radionuclide concentration in leachate that would result in an annual dose of 100 mrem to the landfill worker. These threshold concentrations are several orders of magnitude greater than those measured in the leachate and in all cases are greater than the concentrations in Table 3 of OAR 345-0050-0035 for soluble forms of the radionuclide. In fact, if the leachate concentrations were at the Table 3 limits for the soluble form, the annual dose to the landfill worker after a 50-year buildup time would be 5.8 mrem1.

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Table 18. Radionuclide Concentrations in Leachate Water (pCi L–1) That Would Result in an Annual Effective Dose of 100 mrem to a Landfill Worker Radionuclide Concentration (pCi L-1) U-238 6.35E+05 U-234 1.03E+07 Th-230 1.49E+06 Ra-226 1.81E+04 Pb-210 3.49E+05 Th-232 1.36E+06 Ra-228 6.31E+04 Th-228 3.11E+04 U-235 3.50E+05 K-40 2.74E+05

7. Conclusions

Radiological impacts from leachate management practices at CWMNW are extremely low and do not suggest any changes are necessary to the current leachate management methods. The maximum annual effective dose to a landfill worker who was assumed to spend 30 minutes per day on the landfill surface for 250 days per year from these practices was 0.22 mrem. Annual effective doses to CWMNW employees who work at the laboratory south of the L-14 landfill and the nearest resident were less than 0.005 mrem. The dose to a landfill worker from the disposal of flocked solids and carbon filter media from the on-site water treatment facility (WWT-1) was extremely low at 0.001 mrem. This exposure scenario is extremely unlikely in that it assumes all leachate from the landfill is treated through the on-site water treatment plant when in reality, only a small fraction of the leachate is treated by the system as most of it is applied to the landfill surface as dust control. In all cases the calculated effective doses are substantially less than the 25 mrem yr–1 dose limit recommended by the American National Standards Institute (ANSI 2009) for unrestricted release of soils from land containing TENORM materials, and well below the 100 mrem yr-1 public dose limit set by the Nuclear Regulatory Commission in 10 CFR §20.1301.

Analysis of the Leachate Management Practices 35 at CWMNW

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