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R A C T e c h n i c a l M e m o r a n d u m N o . 5 - C W M N W A r l i n g t o n - 2 0 2 0
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
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R I S K A S S E S S M E N T C O R P O R A T I O N
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
2 CWMNW Arlington, OR
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.
R I S K A S S E S S M E N T C O R P O R A T I O N
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2 CWMNW Arlington, OR
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.
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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.
4 CWMNW Arlington, OR
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.
Wastewater Treatment
Plant
Evaporation Pond-A and
Pond-B
Leachate Spray System - Primary
L-14 cell 1 L-14 cell 2 L-14 cell 3 L-14 cell 4
Sump Sump Sump Sump
Water only sent to wastewater treatment plant during periods of rain/snow and snow cover on landfill Solids/Filter media
sent to landfill
Storage Tank
Chemical flocculants added, particulates
settle to base
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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
6 CWMNW Arlington, OR
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)
Radionuclide
L-14 cell 1 L-14 cell 2 L-14 cell 3 L-14 cell 4 Average
OAR 345-
050-0025
Table 1
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.
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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.
8 CWMNW Arlington, OR
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
Li L
i
C VR
A= (1)
where
Ri = leachate application rate for radionuclide i (pCi m–2 yr–1)
CLi = concentration in the leachate for radionuclide i (pCi m–3)
VL = annual volume of leachate (3,418.9 m3 yr–1)
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
LV
IA
= (2)
1 cm
Percolation (kp)
Aqueous-phaseLeaching (k2)
Inhalation from resuspensionSoil ingestion
External Exp
osu
reSprayer
5 cm
15 cm
Aqueous-phaseleaching (k3)
Q1
Q2
Q3
Leachate to soil (R)
Fraction of spray to air (inhalation)
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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
Cell
Volume
(gal yr–1)a
Volume
(m3 yr–1)
Area of cell
(m2)b
Infiltration
(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.
10 CWMNW Arlington, OR
( )
( )
( )
( )
( )
,
,
,
, 1 , ,
,
, 1 , 1 , ,
,
, 1 1,
,
, 1 , , 1 1,
,
, 1
1, 1
1, 2
1, 3
1, 1
1, 2
i j
i p i i j
i j
p i j i j i i j
i j
i j i j i j i i j
i j
i p i i j i i j
i j
p i j i j i i j i i j
i j
i j
dQR k Q i j
dt
dQk Q k Q i j
dt
dQk Q k Q i j
dt
dQR k Q Q i j
dt
dQk Q k Q Q i j
dt
dQk Q
dt
−
− −
− −
− − −
−
= − + = =
= − + = =
= − + = =
= − + + =
= − + + =
= ( ), 1 , , 1 1, 1, 3i j i j i i j i i jk Q Q i j − − −− + + =
(3)
where
Ri = leachate application rate for radionuclide decay chain member i (atoms m–2 yr–1)
kp = percolation rate constant (0.6769 yr–1)
Qi,j = radionuclide inventory for radionuclide decay chain member i and layer j (atoms
m–2)
ki,j = leach rate constant for radionuclide decay chain member i and layer j (yr–1)
Equations are written in terms of atoms because decay chains are involved. The conversion
from activity to atoms is given by
0.037 dps/pCiactivity
atoms
= (4)
where
Qactivity = radionuclide activity (pCi)
= decay rate constant (s–1)
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
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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)
,
1i j
di bj
Ik
KT
= +
(5)
where
I = water infiltration rate (0.0588 m yr–1)
= moisture content (m3 m–3)
Kdi = soil-water partitioning coefficient for radionuclide i (cm3 g–1)
Tj = thickness of layer j (m)
b = bulk density (g cm3).
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
,
,
1
i j
i j j
b k
k
QC
T=
=
(6)
where Qi,j is the inventory has been converted to activity (pCi m–2) 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)
12 CWMNW Arlington, OR
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
layer (yr–1)
0.6769 NRC (1975) as cited in Peterson (1983)
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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
(m yr–1)
387.2 Carsel and Parrish (1988)
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
(s m–3)
9.91E-6 Calculated with AERMOD
Dispersion factor, nearest resident
(s m–3)
2.78E-7 Calculated with AERMOD
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)
Element
Sand
(mL g–1)a
Loam
(mL g–1)a
Clay
(mL g–1)a
RESRAD
(mL g–1)b
Hanford
(mL g–1)c
INL
(mL g–1)d
GM
(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
14 CWMNW Arlington, OR
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
,1i i iDIh C MLF BR EF DCIh= (7)
where
DIhi = inhalation dose for radionuclide i (mrem per year)
MLF = mass loading factor (1×10–4 g m–3)
BR = breathing rate (m3 d–1)
EF = exposure frequency (d yr–1)
DCIhi = inhalation dose coefficient for radionuclide i (mrem pCi–1)
Ci,1 = concentration for radionuclide i in the surface layer (pCi g–1)
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
,1i i iDIg C INGs EF DCIg= (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)
DCIgi = inhalation dose coefficient for radionuclide i (mrem pCi–1)
Ci,1 = concentration for radionuclide i in the surface layer (pCi g–1)
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
, ,
1
i i j i j
j
DEx C ET DCEx=
= (9)
where
DExi = external exposure dose for radionuclide i (mrem per year)
ET = exposure time (s yr–1)
DCExi,j = external dose coefficient for radionuclide i and layer interval j (mrem-g pCi–1 s–1)
Ci,j = concentration for radionuclide i in layer interval j (pCi g–1)
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
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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
( ), ,1k
i k i
LF
DFC C MLF
DF= (10)
where
Ci,k = annual average air concentration of radionuclide i at receptor k (pCi m–3)
DFLF = dispersion factor in the center of the landfill source (4.21×10–5 s m–3)
DFk = dispersion factor at receptor k (9.91×10–6 s m–3 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
,Li L
i k k i
C VDIn ARF RF DF BR EF DCIh
CF= (11)
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
16 CWMNW Arlington, OR
Parameter Value Reference and Comments
Soil ingestion rate
(mg d–1)
20.6 Based on an 8-hr ingestion rate of 330 mg d–1 (EPA
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
Radionuclide
0-1 cm
Sv m3
(Bq-s)–1
0-5 cm
Sv m3
(Bq-s)–1
0-15 cm
Sv m3
(Bq-s)–1 Fraction
0-1 cm
mrem-g
(pCi-s)–1
0-5 cm
mrem-g
(pCi-s)–1
0-15 cm
mrem-g
(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
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Radionuclide
0-1 cm
Sv m3
(Bq-s)–1
0-5 cm
Sv m3
(Bq-s)–1
0-15 cm
Sv m3
(Bq-s)–1 Fraction
0-1 cm
mrem-g
(pCi-s)–1
0-5 cm
mrem-g
(pCi-s)–1
0-15 cm
mrem-g
(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.
-1
3 -1
(mrem-g [pCi-s] )
(Sv-m [Bq-s] ) 3700 mrem/pCi per Sv/Bq
i
i b
DCEx
DCEx
=
(12)
where
b = bulk density (1.76×106 g m–3)
18 CWMNW Arlington, OR
Table 7. Inhalation and Ingestion Dose Coefficients from DOE-STD-1196 (DOE 2011) as
Presented in RESRAD V2.7 (Yu et al., 2016)
Radionuclide
Inhalation
(mrem pCi–1)
Inhalation Solubility
Typea
Ingestion
(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
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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.
20 CWMNW Arlington, OR
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
( )4.1
3.1
2
2.20016.0
=MC
U
kE (13)
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,
kg
g 1000= CMEQ (14)
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
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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
(15)
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
(16)
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
( )0 0 0 0
0
0
10 1kt ktQ Q Q Q
TIC e dt eV V K VK VK
− −
= = − = − − =
(17)
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.
L =
Vload
H load
+ l
V = L2 Hmc -H load( )
(18)
where
Vload = the volume of the load (~9 m3 per load, 2 loads per year)
Hload = height of the load after disposal (m)
K =U
L
dQ
dt= -KQ
Q(t) =Qoe-Kt
22 CWMNW Arlington, OR
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
1
n
j j
j
DINH IR TIC DCIh=
= (19)
where
DINH = the inhalation effective dose for a disposal of flocked solids and carbon filter
media (mrem)
IR = inhalation rate (m3 s–1)
TICj = time-integrated concentration for radionuclide j (pCi-s m–3)
DCIhj = inhalation effective dose coefficient for radionuclide j (mrem pCi–1)
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
𝐷𝑖𝑛𝑔 =330 mg
day×
1 g
1000 mg×
1 day
8 hours× 0.12 hours × ∑ CA𝑗𝐷𝐶𝐼𝑔𝑗
𝑛𝑗=1 (20)
where
Ding = effective dose from ingestion (mrem)
CAj = weighted average concentration in TENORM for radionuclide j (pCi g–1)
DCIgj = ingestion effective dose coefficient for radionuclide j (mrem pCi–1).
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.
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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
disposal (m3)
Vload 3.0 Calculated
Bulk density (kg m-3) 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
(m)
L 10.7 Calculated from Equation (6)
Air mixing volume (m3) V 22.2 Calculated from Equation (6)
Distance to current nearest
offsite resident (m)
x 3,261 Distance to nearest current
resident estimated from Google
Earth imagery
Removal rate constant (s-1) K 0.5 Calculated using Equation (3)
Emission rate (kg released
to air per load)
E 1.19E-03 Calculated using Equation 1
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
DFETDRD = (21)
where
DR = the dose rate on contact with the material
ET = the exposure time (hours)
DF = distance factor (unitless).
24 CWMNW Arlington, OR
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
from unshielded wastea 9.80E-05 mrem hr-1
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
= t
t
tbt xD
DECJ
tanh104 (22)
where
Jt = flux from the surface of waste layer in the disposal cell (pCi m–2 s–1)
C = Ra-226 concentration in flocked solids and carbon filter media (pCi g–1)
Dt = radon diffusion coefficient in flocked solids and carbon filter media (m2 s–1)
= radon decay constant (2.1×10–6 s–1)
b = bulk density of waste (g cm-3)
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
( )( ) ( )( )
( )( )
−=
−−=
==
−++=
−
−
−
sb
i
iii
ii
xbttctttct
xbt
c
MPm
mkDa
tciDb
exbaaxbaa
eJJ
cc
cc
1110
11
or ,
tanh1tanh1
2
2
2
2
(23)
where
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Jc = radon flux from the disposal cell surface (pCi m–2 s–1)
s = particle density (g cm–3)
= 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
( ) 524exp07.0 mmmDi −−−= (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)
CBAEEC 379.0516.0105.0 ++= (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
170 hours
mrem760
WLM
WLM WL
D WLM
=
=
(26)
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
26 CWMNW Arlington, OR
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
Bulk density, waste (g cm-3) b 1.76 Geosyntec Consultants (2020)
Bulk density, cover (g cm-3) 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
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R I S K A S S E S S M E N T C O R P O R A T I O N
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 0.0980393.0 10 10 u
air waterC C −= (27)
where
Cair = concentration in air (pCi m–3)
Cwater = solute concentration in water (pCi m–3)
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
28 CWMNW Arlington, OR
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
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R I S K A S S E S S M E N T C O R P O R A T I O N
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.
30 CWMNW Arlington, OR
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
Radionuclide Inhalation Ingestion External Total
(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
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R I S K A S S E S S M E N T C O R P O R A T I O N
Radionuclide Inhalation Ingestion External Total
(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
Radionuclide Laboratory Worker Nearest Resident
(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
32 CWMNW Arlington, OR
Radionuclide Laboratory Worker Nearest Resident
(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
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R I S K A S S E S S M E N T C O R P O R A T I O N
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
Pathway
Dose to Landfill Worker Dose to Nearest Current Resident
(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
Pathway Radon Concentration
(pCi L-1) WLM Dose (mrem)
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.
1 This calculation assumes U-235 is at the natural activity abundance of 2.25% relative to U-238
and Th-228 is in secular equilibrium with Ra-228.
34 CWMNW Arlington, OR
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
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R I S K A S S E S S M E N T C O R P O R A T I O N
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