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Home > About NRC > How We Regulate > Rulemaking > Potential Rulemakings > Unique Waste Streams 

> Background Information on Depleted Uranium Background Information on Depleted Uranium The 

uranium fuel cycle begins by extracting and milling natural uranium ore to produce "yellow cake," a 

varying mixture of uranium oxides. Low-grade natural ores contain about 0.05 to 0.3% by weight of 

uranium oxide while high-grade natural ores can contain up to 70% by weight of uranium oxide. 

Uranium found in natural ores contains two principle isotopes – uranium-238 (99.3%) and uranium-235 

(0.7%). The uranium is enriched in uranium-235 before being made into nuclear fuel. Uranium 

enrichment processes generate a product consisting of 3 to 5 percent uranium-235 for use as nuclear 

fuel and a product of depleted uranium (about 0.3 percent U-235). The depleted uranium has some 

commercial applications including counterweights and antitank armaments. However, the commercial 

demand for depleted uranium is currently much less than the amounts generated. For instance, the U.S. 

Department of Energy (DOE) has about 750,000 metric tons of depleted uranium in storage. Under the 

U.S. Enrichment Corporation Privatization Act, DOE is required to accept depleted uranium from a U.S. 

Nuclear Regulatory Commission (NRC) licensed uranium enrichment facility if the depleted uranium is 

determined to be low-level radioactive waste. If the depleted uranium has no commercial use, the 

licensee can transfer the material to DOE or dispose of it at a commercial disposal site if it meets the 

disposal site’s requirements.

For more information on Depleted Uranium, see below: 

•Uses

•Health Effects

•Toxicological and Radiological Concerns 

•Current Issues 

Uses DU is used in the manufacturing of ammunitions used to pierce armor plating, such as those found 

on tanks, in missile nose cones and as a component of tank armor. Armor made of depleted uranium is 

much more resistant to penetration by conventional anti-armor ammunitions than conventional hard 

rolled steel armor plate. 

Armor piercing ammunitions are generally referred to as "kinetic energy penetrators". DU is preferred to 

other metals, because of its high density, its pyrophoric nature (DU self-ignites when exposed to 

temperatures of 600° to 700° and high pressures), and its property of becoming sharper, through 

adiabatic shearing, as it penetrates armor plating . On impact with targets, DU penetrators ignite, 

breaking up in fragments, and forming an aerosol of particles ("DU dust") whose size depends on the 

angle of the impact, the velocity of the penetrator, and the temperature. These fine dust particles, can 
catch fire spontaneously in air. Small pieces may ignite in a fire and burn, but tests have shown that 

large pieces, like the penetrators used in anti-tank weapons, or in aircraft balance weights, will not 

normally ignite in a fire. 

Health effectsUranium is introduced into the body mainly through ingestion of food and water and 

inhalation of air. When inhaled, uranium is attached to particles of different sizes. The size of the 

uranium aerosols and the solubility of the uranium compounds in the lungs and gut influence the 

transport of uranium inside the body. Coarse particles are caught in the upper part of the respiratory 

system (nose, sinuses, and upper part of the lungs) from where they are exhaled or transferred to the 

throat and then swallowed. Fine particles reach the lower part of the lungs (alveolar region). If the 

uranium compounds are not easily soluble, the uranium aerosols will tend to remain in the lungs for a 

longer period of time (up to 16 years), and deliver most of the radiation dose to the lungs. They will 

gradually dissolve and be transported into the blood stream. For more soluble compounds, uranium is 

absorbed more quickly from the lungs into the blood stream. About 10% of it will initially concentrate in 

the kidneys. 

Most of the uranium ingested is excreted in feces within a few days and never reaches the blood stream. 

The remaining fraction will be transferred into the blood stream. Most of the uranium in the blood 

stream is excreted through urine in a few days, but a small fraction remains in the kidneys and bones 

and other soft tissue. 

 

Toxicological and Radiological concernsIn sufficient amounts, uranium that is ingested or inhaled can be 

harmful because of its chemical toxicity. Like mercury, cadmium, and other heavy-metal ions, excess 

uranyl ions depress renal function (i.e., affect the kidneys). High concentrations in the kidney can cause 

damage and, in extreme cases, renal failure. The general medical and scientific consensus is that in cases 

of high intake, uranium is likely to become a chemical toxicology problem before it is a radiological 

problem. Since uranium is mildly radioactive, once inside the body it also irradiates the organs, but the 

primary health effect is associated with its chemical action on body functions. 

In many countries, current occupational exposure limits for soluble uranium compounds are related to a 

maximum concentration of 3 μg uranium per gram of kidney tissue. Any effects caused by exposure of 
the kidneys at these levels are considered to be minor and transient. Current practices, based on these 

limits, appear to protect workers in the uranium industry adequately. In order to ensure that this kidney 

concentration is not exceeded, legislation restricts long term (8 hour) workplace air concentrations of 

soluble uranium to 0.2 mg per cubic meter and short term (15 minute) to 0.6 mg per cubic meter. 

Like any radioactive material, there is a risk of developing cancer from exposure to radiation emitted by 

natural and depleted uranium. The annual dose limit set by the IAEA for a member of the public is 1 

mSv, while the corresponding limit for a radiation worker is 20 mSv. The additional risk of fatal cancer 

associated with a dose of 1 mSv is assumed to be about 1 in 20,000. This small increase in lifetime risk 

should be considered in light of the risk of 1 in 5 that everyone has of developing a fatal cancer. It must 

also be noted that cancer may not become apparent until many years after exposure to a radioactive 

material. 

Current IssuesExisting NRC regulations at 10 CFR 61.55, "Waste Classification," specify criteria for 

determining the classification of low-level radioactive waste for land disposal at a near-surface facility. 

The original development of 10 CFR 61.55 did not explicitly consider the impacts resulting from the 

disposal of unique waste streams such as significant quantities of depleted uranium from the operation 

of a commercial uranium enrichment facility. When 10 CFR Part 61, "Licensing Requirements for Land 

Disposal of Radioactive Waste," was initially developed, there were no commercial facilities generating 

significant quantities of depleted uranium waste streams. As a result the analysis only considered the 

types of uranium-bearing waste streams being typically disposed of by licensees at the time. 

Additionally, the nature of the radiological hazards associated with DU presents challenges to the 

estimation of long-term effects from its disposal – namely that its radiological hazard gradually increases 

due to the ingrowth of decay products, eventually peaking after 1 million years, rather than decreasing 

significantly over a few hundred years like that of typical LLW. 

The U.S. Nuclear Regulatory Commission (NRC) staff has identified several key issues for initial 

discussion with stakeholders on disposal of DU. These include defining key regulatory terms such as 

unique waste streams and significant quantities of depleted uranium as well as technical parameters of 

a site-specific analysis including a time period of performance, appropriate exposure scenarios for 

protection of the public and individuals from inadvertent intrusion. The NRC staff is also soliciting 

stakeholder views on technical issues for a site-specific analysis of near-surface disposal of significant 

quantities of depleted uranium. These technical issues include appropriate considerations for depleted 

uranium waste form(s), uranium geochemistry, and radon migration and exposure. These issues arose 

from the results of the NRC staff’s technical analysis (SECY-08-0147) that was submitted to the 

Commission on October 7, 2008, in response to Commission Order CLI-05-20 regarding depleted 
uranium. Given those issues, the Commission's related Staff Requirements Memorandum (SRM-SECY08-0147), dated March 18, 2009, instructed the staff to begin engagement with stakeholders and 

interested parties to initiate development of the technical basis for possible revision of the 10 CFR Part 

61, "Licensing Requirements for Land Disposal of Radioactive Waste." Toward that end, the staff has 

scheduled public workshops to discuss the benefits and impacts of revising 10 CFR Part 61. In so doing, 

the staff hopes to identify potential conflicts and gain an understanding of any unintended 

consequences that may result from drafting and implementing related changes to the NRC's existing 

regulations. 
http://www.nrc.gov/reading-rm/doc-collections/commission/secys/2008/secy2008-

147/enclosure1.pdf 

ANALYSIS OF DEPLETED URANIUM DISPOSAL OVERVIEW: 

A screening model has been developed by staff of the performance assessment branch in the 

Division of Waste Management and Environmental Protection to evaluate the risk and 

uncertainties of depleted uranium (DU) disposal as low-level waste (LLW) with near-surface 

disposal at a generic site. The model was developed to understand the impacts of key variables 

on the risks from disposing of DU in near-surface disposal, such that staff could respond to 

Commission direction to consider whether the quantities of DU in the waste stream from 

uranium enrichment facilities warrant amending the waste classification tables in 10 CFR Part 61 

(Part 61). The model was developed to evaluate the radiological risk to potential future residents 

and intruders (acute or chronic exposures) near or on the land overlying a hypothetical disposal 

facility for the large quantities of DU anticipated to be disposed of as a result of fuel enrichment 

facility operations. The model was designed to provide the user flexibility in evaluating different 

waste types and forms, disposal configurations, performance periods, institutional control 

periods, pathways, and scenarios. 

The model was constructed with the dynamic simulation software package GoldSim®, 

developed by GoldSim Technology Group of Issaquah, WA. Goldsim is a Monte Carlo 

simulation software solution for dynamically modeling complex systems in business, 

engineering and science. GoldSim is used for decision and risk analysis by simulating future 

performance while quantitatively representing the uncertainty and risks inherent in all complex 

systems. GoldSim has been used by U.S. Nuclear Regulatory Commission (NRC) staff to 

iskinform reviews of U.S. Department of Energy (DOE) performance assessments (Esh, 2002; 

Esh, 2006). GoldSim is used by over 30 organizations in the field of radioactive waste 

management. A component or modular approach can be used in GoldSim to build a performance 

assessment model, which is the approach used in this analysis. Main submodels include 

inventory, source term, infiltration, radon, groundwater transport, and biosphere. 

Submodels use deterministic and probabilistic input values or distributions. 

The model was used to understand the impacts of key variables on the risks from disposing of 

DU in near-surface disposal. Key variables evaluated were: disposal configurations, performance 

periods, institutional control periods, waste forms, site conditions, pathways, and scenarios. 

Calculations were performed probabilistically to represent the impact of variability and 

uncertainty on the results. The analysis methodology in the current assessment was consistent 

with the technical analysis methodology used for the development of the environmental impact 

statements supporting Part 61. This approach allowed constraints to be identified for the safe 

disposal of large quantities of DU in near-surface disposal. Because there were a wide range of 

variables considered, summary conclusions are not absolute; a sitespecific analysis may 

demonstrate compliance with the performance objectives when the summary conclusions found 

below indicate otherwise. However, the properties and characteristics of DU present constraints 

on approaches for disposal. The summary conclusions provide the technical framework for 

policy decisions. The main technical observations are: 
• Depleted uranium has some characteristics that are dissimilar from commercial LLW. 

– A large percentage of the activity is associated with very long-lived radionuclides. 

ii 

– Radioactive decay results in increasing hazard with time until after 1 million years, as 

a result of increasing concentrations (and higher mobility) of decay products. 

– In-growth of significant quantities of a daughter in gaseous form (222Rn) 

• Estimated risks are sensitive to the performance period. 

• Estimated risk from radon is sensitive to the disposal depth. 

• Radon fluxes to the environment are very sensitive to the long-term moisture state of the 

system.

• Large uncertainties (and little available data) associated with some transfer factors for uranium 

daughter products. 

• Estimated disposal facility performance is strongly dependent on site-specific hydrologic and 

geochemical conditions. 

• Radon is limiting at arid sites and for shallow disposal. 

• The groundwater pathway is limiting at humid sites. 

• Grouting of the waste may improve the likelihood of an arid site meeting the performance 

objectives with respect to radon; however, grout may enhance the mobility of uranium in the 

groundwater pathway after the grout degrades. 

The summary conclusions from the technical analysis are: 

• Near-surface disposal (i.e., less than 30 meters [m], as defined in Part 61) may be appropriate 

for large quantities of DU under certain conditions. However, unfavorable site conditions can 

result in the performance objectives not being met. Examples of unfavorable conditions include 

shallow disposal (< 3 m depth) and humid sites with a potable groundwater pathway. 

• Because of the in-growth of radon and other daughter products, periods of performance of 

1,000 years or less result in a significant truncation of estimated risk. 

• Shallow disposal (< 3m deep) is likely to not be appropriate for large quantities of DU, 

regardless of site conditions. Shallow disposal may be possible if robust intruder barriers, 

excluding the possible excavation of DU, and a robust radon barrier that can effectively limit 

radon fluxes over the period of performance are installed, and their performance is justified. 

Small quantities (1 – 10 metric tons) could be disposed of at shallow depths. 

• Depleted uranium can be disposed of under arid conditions and meet the Part 61 performance 

objectives for 1,000 to 1 million year performance periods, if the waste disposal depth is large,

or robust barriers are in place to mitigate radon.

• Disposal under humid conditions with viable water pathways is probably not appropriate for 

large quantities of DU. 

iii 
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R-1 8
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R-23
APPENDIX A
A.1 GENERIC ACCIDENT METHODOLOGY

A.1.1 Introduction 

A generic methodology was developed to estimate radiation doses from accidents involving 

Nuclear Regulatory Commission (NRC)-exempted products and materials. The methodology 

includes the following accident exposure scenarios: (1) fires involving the release of radioactive 

materials from all types of products, (2) mechanical resuspension of radioactive material during 

cleanup after a fire, (3) spills of radioactive materials in liquid or powder form, and (4) crushing 

of glass tubes containing radioactive gases. 

For these selected accident exposure scenarios, radionuclide-specific dose-to-source ratios 

(DSRs) are calculated. The DSRs give the radiation dose associated with a unit quantity of the 

radioactive material at risk in an accident. In the derivation of the DSRs, the methods and 

parameter values used in prior assessments addressing similar issues were used when 

applicable. The methods and parameters used in the development of this generic methodology 

are discussed and used in the derivation of the DSRs for a variety of radionuclides in many 

exempted products. Because of the broad range of exemptions, it has been necessary to 

estimate radiation doses for some exempted products or materials on a case-by-case basis 

when the DSR methodology is not applicable. In some instances, radiation doses have also 

been estimated on a case-by-case basis using better data when available for a specific 

accident exposure scenario and product of interest. 

A.1.2 Airborne Concentrations 

Airborne concentrations of radioactive materials during an accident and cleanup following an 

accident are estimated using two equations. For the instantaneous release of radioactive 

materials during an accident, the average airborne concentration (microcurie (4Ci)/m3 ) is given 

by 

C = Q(1 - e-kt)Vk (1 

where Q = amount of radioactive material (gCi) released at t = 0, 

k = ventilation rate (h-'), 

t = time over which C is averaged (h), and 

V = volume of air into which material is dispersed (m3 ). 

For cleanup following an accident involving fire, the average airborne concentration (gCi/m3

) is 

given by 

C = K x S, (2) 

where S = level of contamination on a surface (4Ci/m2), and 

K = empirically determined factor for mechanical resuspension of respirable size 

particles (m-').

A.1-1
A-1.3 Resuspension Factors

Particle resuspension from a contaminated surface has been reviewed by Healy (1980), Schmel 

(1980), and Nicholson (1988). The factor used here for mechanical resuspension stresses on a 

contaminated surface is lx10-5 m-. The use of this value is supported by practices used in 

both radiation protection (Carter, 1983) and transportation (Humphries and Dodd, 1989). 

A.1.4 Release Fractions 

Release fractions are used to estimate the release of airborne radioactive material during a spill 

or fire (see Table A.1.1). Hence, the amount of radioactive material (/.Ci) released into air in 

Equation 1 is given by 

Q = RF x A , (3) 

where A = total amount of radioactive material involved in the spill or fire (pUCi), and 

RF = fraction of radioactive material released as respirable size particles (unfitness). 

In the case of a fire, a release fraction of 0.1% is assumed for most materials; however, for 

gases, a value of 100% is assumed, and for solid materials in protective devices, a value of 

0.01% is assumed. If better data are available in a particular case (e.g., a lower release of 

uranium in glassware or thorium in alloys in a fire), then those better data have been used. The 

practices in transportation (International Atomic Energy Agency (IAEA) SS No. 7) and the 

release fractions in 10 CFR 30.72 provide general support for the use of 0.1% in many cases. 

In the case of a spill, a release fraction of 0.1% is assumed, provided only small amounts of a 

simple liquid or solid and temperatures of less than 100'C are involved (Environmental 

Protection Agency (EPA) 520/1-89-001). If the spill involves higher temperatures and either 

volatile solids or flammable liquids, the use of other release factors should be considered 

(EPA 520/1-89-001; DOE-STD-1027-92). If the spill involves a large amount of liquid, the 

use of a simple release factor may not be appropriate (Martin Marietta Energy Systems 

(MMES), 1992). 

A.1.5 Inhalation Intakes 

During a spill or fire, an individual's intake from inhalation of an airborne radioactive material 

(/.Ci) is given by 

I = Cx BRx t, (4) 

where t = time of exposure (h), 

BR = breathing rate (m3

/h), and 

C = average concentration of the airborne radioactive material (/uCi/m 3) over 

the time, t.

A.1-2
Some enclosure volumes and ventilation rates used throughout this report are summarized in 

Table A.1.2. A ventilation rate of 1 volume change per hour for a house is within the expected 

range of 0.25 to 2 from a recent report by Koontz and Rector (EPA Contract No. 68-D9-01 66). 

A ventilation rate of 1 volume change per hour for a large warehouse is within the expected 

range of 1 to 4 from a report by the American Society of Heating, Refrigeration, and Air 

Condition Engineers (ASHRAE) (1991). Table A.1.2 also includes data for a laboratory-type 

room that is thought to be typical of those found at many industrial facilities and educational 

institutions such as colleges and high schools. The enclosure volume and ventilation rate 

are based on the approximate size of a 20-student laboratory or 30-student classroom 

(ASHRAE, 1993). A ventilation rate of 6 volume changes per hour for a laboratory-type room is 

within the expected range of 6 to 12 for a chemical laboratory at an industrial or commercial 

facility (ASHRAE, 1995; National Research Council, 1995). 

The breathing rate assumed for general use throughout this report is the daily average rate of 

approximately 0.9 m3/h (i.e., 22 m3/day or 8000 m3/yr). However, a breathing rate for light 

exercise of 1.2 m3

/h has been used for transportation accidents involving fire (IAEA SS No. 7), and this value has been applied in all accident scenarios. In the case of a fire inside an 

enclosure, it is assumed that a firefighter wears a supplied-air respirator that is operated in a 

pressure-demand or other positive-pressure mode and that provides a protection factor of 1000 

(NIOSH, 1990). 

A.1.6 Ingestion Intakes 

For ingestion of materials in powder or liquid form, the materials presumably would not be 

deliberately ingested during normal use but could be transferred to the hands during a spill and 

cleanup following a spill. Once the material is transferred to an individual's hands or skin, it can 

be ingested directly or absorbed through wetted skin in the case of tritiated water. 

For direct ingestion of liquids or powders, assumptions made in this report are consistent with 

those used by the IAEA in assessing transportation accidents (IAEA SS No. 7). It is assumed, 

first, that 10% of the available material would be deposited somewhere on the skin of an 

individual and, second, that 0.1% of this deposited material would be ingested before bathing 

removed the material from the body. Thus, the direct ingestion intake, /, is estimated to be 

lx1 0-4 times A, where A is the total amount of available material at risk in the accident. 

For absorption of tritiated water through wetted skin, the equations of Osborne (1966) could be 

applied. However, it is assumed that 10% of the tritiated water is spilled somewhere on the skin 

of the individual and is completely absorbed through the skin before bathing. Thus, the intake 

of tritiated water through the wetted skin is estimated to be 10% times A or 1 0x1 01 A, where A 

is the total amount of tritiated water at risk in the accident. 

A.1.7 Radiation Dose Estimation 

Radiation doses are estimated using the effective dose equivalent (EDE) based on the 

International Commission on Radiological Protection (ICRP) 26 approach that radiation risk 

should be the same, whether the whole body is irradiated uniformly or several organs receive all 

of the radiation dose. The ICRP 26 approach replaces the critical organ concept that was used 

for many years. The EDE is the sum of the radiation doses to each organ, after the dose

A.1-3
equivalent for each organ is multiplied by an organ weighting factor based on somatic health 

risk estimates from many studies (ICRP 26). 

In addition to the radiation doses resulting from inhalation or ingestion, it is also possible to 

receive a radiation dose from submersion in airborne radioactive materials and from 

resuspension of any radioactive materials during cleanup following a fire. These potential 

exposure pathways are all considered in the following discussions. In the case of submersion 

that may result in a significant radiation dose being delivered to the skin, a skin weighting factor 

of 0.01 is also used in estimating the EDE. This recommendation was made by the ICRP 

following ICRP 26 (Kocher and Eckerman, 1988) and incorporated in the later 

recommendations of ICRP Publication 60. 

The EDE (rem) to an individual from inhalation and ingestion of a radioactive material is given 

by 

H = DCFx I, (5) 

where I = intake of an individual by inhalation or ingestion (guCi), and 

DCF= dose conversion factor for the 50-year committed EDE from inhalation or 

ingestion (rem//MCi) (see Sections 2.1 and 3.1, and EPA-520/1--88-020). 

The EDE (rem) to an individual from submersion in airborne radioactive material is given by 

H = DCFx Cx t, (6) 

where t = time of an individual's exposure (h), 

C = average concentration of the airborne material (UCi/m 3

) over the time, t, and 

DCF = dose conversion factor for air submersion (rem-m3

/jpCi-h) (see Sections 2.1 

and 3.1, and EPA-402-R-93-081). The skin dose component may be 

included using a 0.01 weighting factor. 

The DSRs in the following sections of this report are the EDE (for inhalation, ingestion and 

submersion), calculated per uCi of a parent byproduct material (e.g., 00Co or 13 7

Cs) or mg of a 

parent source material (e.g., natural thorium). However, when calculating the DSRs, it is 

necessary to convert the mass of a parent source material (mg) to activity (,gCi). For example, 

the mass of natural thorium is due almost entirely to 232

Th, which has a specific activity of 232

Th 

of 0.0403 terabecquerel (TBq/g (1.09x10- 7 Ci/g) (see Table 3.1.3 of Section 3.1), and the 

activity of the 232Th in 1 mg of natural thorium is 4.03 Bq (1.09x10-V /Ci). If 20 years have 

elapsed since the natural thorium was chemically separated (see Table 3.1.5 of Section 3.1), 

the total activity in the 1-mg source is as follows: 4.03 Bq (1.09x10-4 /Ci) of 232Th; 3.67 Bq 

(9.92x10-5 MCi) of 2

8 Ra; 3.51 Bq (9.48x10' /MCi) of "8Th and 224Ra; and some additional 

activity from a number of short half-life products of decay (see Table 3.1.3 of Section 3.1) that 

do not contribute significantly in most of the dose calculations.

A.1-4
A.1.7.1 Radiation Dose From Inhalation During an Accident Involving Fire

For completeness, three types of fires are considered: (1) warehouses belonging to 

manufacturers or distributors that may contain large numbers of exempted products, (2) transportation accidents in which a few cartons or pallets of exempted products are 

involved, and (3) residences in which only a few exempted products are involved (see Tables A.1.4 through A.1.6). In many of the previous assessments, residential and warehouse 

fires have been considered, but not transportation fires. Transportation fires were included 

here because many exempted products may be shipped to vendors or consumers without being 

stored in large numbers. 

For transportation accidents occurring indoors (i.e., storeroom or cargo-handling bay) or 

outdoors (i.e., transportation vehicles) and involving fire, the IAEA (SS No. 7) estimates that the 

inhalation intake during a 30-minute period following the start of the fire ranges from about 0.01 

to 0.1%. The IAEA recommends the use of an intake factor, I, of 1.Oxl 0- Q for a firefighter or 

bystander in the plume of smoke from the fire. It is assumed that a bystander would not stand 

in the plume of smoke from a fire and that a firefighter who is in the plume from the fire would 

wear a supplied-air respirator. Hence, an inhalation intake factor, /, of 8.6x1 0' Q is used, 

based on the average concentration in the air of a storeroom or cargo-handling bay for 

30 minutes following the start of a fire and the use of a supplied-air respirator with a protection 

factor of 1000 (see Table A.1.3). 

The first column of Table A.1.4 presents the DSRs used in this report for a firefighter at a 

transportation accident involving a fire. The equation for calculating the DSRs (rem/jtCi or 

rem/mg) is obtained using Equations (3) and (4) and is given by 

DSR = 8.6x10-7 DCF, x RF, x Ai, (7) 

where DCF, = dose conversion factor for inhalation of a radionuclide i (rem//Ci), 

RF, = release factor for a radionuclide i (unfitness), and 

A, = activity of each radionuclide i (,uCi) per 1 uCi of a parent byproduct material 

or 1 mg of a parent source material. 

The DSRs developed here are based on a release factor, RF, of 100% for gases (e.g., tritium 

(3

H) and 85Kr) and 0.1% for solids, powders, or liquids (see Table A.1.1). If a solid is contained 

in a protective device (e.g., 241

Am in a smoke detector), the DSRs for inhalation of solids in the 

first column of Table A.1.4 should be reduced by a factor of 10. 

For warehouse and residential fires, the same equation as above is used, except for the 

numerical constant. The numerical constant used for a residential fire was 1.Oxl0-6 and that 

used for a warehouse fire was 1.6x1 0-7. These constants are based on the inhalation intakes, /, given in Table A.1.3 for a ventilation rate of 1 volume change per hour in both a residence 

and a warehouse. It should be noted from the data in Table A.1.3 that a ventilation rate of 1 air 

change per hour will predict the inhalation intake, I, to within a factor of 2 over the wide range of 

ventilation rates expected for residences (EPA Contract No. 68-D9-0166) and warehouses 

(ASHRAE, 1991). Warehouses are usually not air conditioned, but they are often heated and 

ventilated sufficiently to provide a tolerable working situation. The DSRs developed here for

A.1-5
firefighters at warehouse and residential fires are provided in the first column of Tables A.1.5 

and A.1.6, respectively. A blank space in a table means that the exposure pathway is not 

operative for that particular radionuclide (e.g., 8

"Kr is an inert gas that is not absorbed into the 

body via the inhalation pathway). 

A.1.7.2 Radiation Dose From Submersion During an Accident Involving Fire 

For transportation accidents involving fire, the IAEA (SS No. 7) uses an average concentration, 

C, of 1.44x1 0- 3 Q per m3 for the radioactivity in air during the 30-minute exposure time following the start of the fire (see Table A.1.3). The equation for calculating the DSRs (rem//pCi or 

rem/mg) from submersion in contaminated air is obtained using Equations (3) and (6) and is 

given by 

DSR = 7.21x10- 4 hr/mr3 • DCF, x RFj x A,, (8) 

where DCF, = dose conversion factor for submersion in a radionuclide i (rem-m 3

/pUCi-h), 

RF, release factor for a radionuclide i (unfitness), and 

Ai = activity of each radionuclide i (MCi) per 1 /pCi of a parent byproduct material 

or 1 mg of a parent source material. 

The DSRs developed here are based on a release factor, RF, of 100% for gases (e.g., 3

85 H and Kr) and 0.1% for solids, powders, or liquids (see Table A.1.1). If a solid is contained in a 

protective device (e.g., 241

Am in a smoke detector), then the DSRs for submersion in air 

contaminated by solids in Tables A.1.4, A.1.5, and A.1.6 should be reduced by a factor of 10. 

For warehouse and residential fires, the same equation as above is used, except for the 

numerical constant. The numerical constant used for a residential fire was 8.75x1 0-4 h/m 3 and 

that used for a warehouse fire was 1.31 x10-4 h/m 3

. These constants are based on the average concentration, C, given in Table A.1.3 for a ventilation rate of 1 volume change per hour in both 

a residence and a warehouse. It should be noted from the data in Tables A.1.4 through A.1.6 

that submersion may be an important exposure pathway for a firefighter who is wearing a 

supplied-air respirator, which reduces the EDEs for inhalation by a factor of 1000 or more. 

For residential fires, consideration is given to doses to a person escaping the fire or a neighbor trying to rescue a person from a fire. Inhalation doses for a nonfirefighter will be greater than 

for a firefighter who wears a supplied-air respirator and has an individual intake of 1.0x10-6 0 

(see value for a residence with a ventilation rate of 1 volume change per hour in Table A.1.3). If an escape or rescue time of 10 minutes is assumed, then individual intakes for nonfirefighters 

are calculated to be 4.1x10-4 Qm 3, and their inhalation doses would be 410 times greater than 

that estimated for firefighters (i.e., 4.1x10 -4 01.0xl0-6 Q). Thus, the DSRs in the first column 

of Table A.1.6 can be multiplied by 410 to estimate the inhalation dose and the DSRs in the 

second column of Table A.1.6 can be multiplied by 0.39 to estimate the submersion dose to a 

nonfirefighter (i.e., a person escaping from a residential fire or a neighborhood hero trying to 

rescue a person from a residential fire).

A.1-6
A.1.7.3 Radiation Dose From Mechanical Resuspension Following a Fire

For the cleanup following a fire, it is assumed, first, that the mechanically resuspendable activity is 1 % of the total activity involved in the fire and, second, that it is spread uniformly over the 

total floor area of the enclosure involved in the fire (NUREG/CP-0001, Wrixon and Freke). Thus, the level of contamination, S, of a storeroom or cargo-handling bay with a floor area 

of 100 m2 (see Table A.1.2) is estimated to be 1% times A divided by 100 m2 or 1x10-4 A m-2

, where A is the total activity involved in the fire. The equation for calculating the DSRs (rem/4Ci 

or rem/mg) from mechanical resuspension following a fire is obtained using Equations (2), (4), 

and (5) and is given by 

DSR = 1.0x10- 4 m-2Z DCFi x Aix K x BR x t, (9) 

where DCF, = dose conversion factor for inhalation of a radionuclide i (rem/yCi), A = activity of each radionuclide i (12Ci) per 1 uCi of a parent byproduct material 

or 1 mg of a parent source material, 

K = mechanical resuspension factor of lx1 0- m-, 

BR = breathing rate of 1.2 m3

/h, and 

t = exposure time, which is assumed to be an 8-hour work day. 

For mechanical resuspension of radioactive material following a warehouse or residential fire, the same equation is used except for the constant value. The constant values used for a 

residence and warehouse are 5.4x10 5

- m-2 and 1.1x10-5 m-2, respectively. 

As an example of the use of the DSRs in Tables A.1.4 through A.1.6, consider a residential fire that destroys two smoke detectors containing 37 kBq (1 pmCi) each of 241

Am (see Section 2.15). The total amount of 241Am material involved in the fire is 74 kBq (2 /MCi). Because 241Am in 

smoke detectors has a release rate of approximately 10-4 from actual measurements 

(NUREG/CR-0403), the smoke detector acts as a protective device, and the DSRs for inhalation and submersion during the fire should be reduced by a factor of 10 (see Footnote "a" 

to Table A.1.6). The individual EDEs to a firefighter are estimated to be (4.4xl 0' rem/,UCi) x 2 Ci+ by 10 or 0.009 nanosievert (nSv) (0.9 nrem) for inhalation and (9.8x10-' 2 rem/M.Ci) x 

2 MCi + by 10 or 2x10-5 nSv (0.002 nrem) for submersion. Thus, the estimated individual EDE 

to a firefighter is due almost entirely to inhalation of 241

Am from the smoke detectors. 

For the cleanup following a residential fire under the same conditions, the committed EDE to an 

individual from inhalation of resuspended material during a work day of 8 hours is estimated to 

be (2.3x10 8

- rem//Ci) x 2 AzCi, or 0.05 MSv (5 mrem). Because the spread of contamination on 

the floor is considered to be independent of the release of material into the air, the factor of 10 reduction is not applied to the EDEs for the inhalation from resuspension of material on the floor 

of the house. For fire inspectors who spend about 25% of their time inspecting mostly house 

fires during a year (i.e., 62.5 working days), annual individual EDEs are estimated to be 

0.05 MSv (5 Mrem) per day (8 hours) times 62.5 working days (500 hours), or 0.003 mSv 

(0.3 mrem). 

Finally, consider the radiation dose to a person escaping from a fire or a neighbor trying to 

rescue a person from a fire under the same conditions. Their inhalation dose is estimated to be

A.1-7
410 times that to a firefighter (i.e., 410 x 0.9 nrem), or 0.004 /zSv (0.4 /.rem), and their 

submersion dose is estimated to be 0.39 times that to a firefighter (i.e., 0.39 x 2 prem), or 

0.01 pSv (1 prem). 

It should be noted that for the above examples, all dose values less than 1x10-5 mSv 

(<0.001 mrem) would have been reported as "less than values" (i.e., less than 1x10 5 mSv 

(<0.001 mrem)) if they had been included in a specific assessment presented in this report. 

This is consistent with the notation that was used for this report. 

A.1.7.4 Radiation Dose From Spilled Liquids and Powders 

For inhalation following a spill, an individual intake, I, of 1.0x10- 3 Q is used, based on a 

30-minute exposure time in a laboratory-type room with an enclosure volume of 180 m3 and a 

ventilation rate of 6 volume changes per hour (see Table A.1.7). The equation used to 

calculate the DSRs (rem//MCi or rem/mg) for inhalation is obtained using Equations (4) and (5) 

and is given by 

DSR = 1.0x10-3 • DCF, x Ai x RF, (10) 

where DCF, = dose conversion factor for inhalation of a radionuclide i (rem/,UCi), 

A, = activity of each radionuclide i (.Ci) per 1 uCi of a parent byproduct material 

or 1 mg of a parent source material, and 

RF = release factor of 0.1% for spills of liquids or powders. 

It should be noted that the individual intake, /, of 1.Oxl 0-3 Q used in the above equation also 

provides an estimate of the maximum inhalation dose to an individual for longer exposure times 

following the spill (i.e., 1 hour or more) at the expected ventilation rate of 6 volume changes per 

hour or more in a laboratory-type room (see Table A.1.7 and Section A.1.5). 

For ingestion following a spill with the exception of 3H, an individual intake factor, /, of 

1.0xl0-4 A is assumed, where A is the total amount of activity involved in the spill (see 

Section A.1.6). The equation for calculating the DSRs (rem/LLCi or rem/mg) for ingestion of a 

material other than 3H is obtained using Equation (5) and is given by 

DSR = 1.0x10-4 DCFi x Ai , (11) 

where DCF, = dose conversion factor for ingestion (rem/p/Ci), and 

A j= activity of each radionuclide i (I.Ci) per 1 Ci of a parent byproduct material 

or 1 mg of a parent source material. 

For 3H, the same equation as above is used, except the numerical constant is 1.OxlO1 (see 

Section A.1.6). Table A.1.8 presents the DSRs for both ingestion and inhalation following a spill 

of a liquid or powder.

A.1-8
As an example of a dose calculation for a spill, consider a quantity of 100 ,UCi (0.37 MBq) of 14C 

obtained as a liquid in the form of a labeled organic compound (see Section 2.13). By the use 

of Equation (10), the DSR for inhalation is (1.0xl0-3) x (2.1x10-3 rem/pCi) x 1 /Ci x (1.0x10-3 ), 

or 5.7x10-10 mSv/kBq (2.1x10 9 rem//pCi) of "

4C, and by the use of Equation (11), the DSR for 

ingestion is (1.0x10- 4

) x (2.1x10-3 rem//Ci) x 1 pCi, or 5.7x10-' Sv/GBq (2.1x10-7 rem//pCi) of 

1

4

C. Thus, the individual EDEs for the inhalation exposure pathway, the ingestion exposure 

pathway, and both pathways combined for the spill of the 0.37 MBq (100 gCi) of 14C are 

estimated to be 0.002 /Sv (0.2 yrem), 0.2 mSv (20 prem), and 0.2 mSv (20 prem), 

respectively. 

(It should be noted that since these doses are less than 1x10' mSv (<0.001 mrem), they would 

have been reported as "less than values" (i.e., less than 1x10' mSv (<0.001 mrem)) if they had 

been part of an assessment in this report. This is consistent with the notation that was used for 

this report). 

A.1.7.5 Radiation Dose From Crushing of Glass Tubes Containing Radioactive Gases 

Table A.1.9 presents DSRs for crushing of glass tubes containing 3

H in the form of tritiated 

water vapor (HTO) or the noble gases 85

Kr and 220

Rn. The 2°Rn is assumed to come from the 

radioactive decay of natural thorium in a glass tube such as a lamp or fluorescent lamp starter 

(see Section 3.7). The external doses from air submersion in 8

"Kr were calculated using Equations (1), (3), and (6), and the external doses from inhalation of 3H in the form of HTO or 

22Rn and its short-lived decay products were calculated using Equations (1), (3), (4), and (5). The exposure times needed to approach the maximum radiation doses are shorter in 

enclosures with high ventilation rates than in enclosures with low ventilation rates. Hence, the 

exposure times were varied in the calculations to give an indication of how long it would take to 

approach the maximum radiation dose for the various types of enclosures considered in 

Table A.1.9. 

As an example of the use of the DSRs in Table A.1.9, consider a night sight for a gun with three 

gaseous tritium light sources (GTLSs) containing approximately 66.6 MBq (18 mCi) of 3H each 

(see Section 2.14). The tritium in the GTLSs is assumed to be 99% elemental 3

H and only 1% 

HTO, so that the amount of HTO in the GTLS is 6.66 MBq (180,pCi). If one of the GTLSs is 

crushed in a home with a volume of 450 m3 and a ventilation rate of 1 volume change per hour, 

then the committed EDE to an individual in the house from inhalation of HTO over the next 

8 hours would be 2.6x10-7 rem/pCi times 180 pCi, or approximately 5x10-4 mSv (0.05 mrem). 

The elemental 3H can be ignored because it contributes insignificantly to the radiation dose 

received by the individual (see ICRP 68, Table C.1). 

A.1.8 Summary 

A generic methodology was developed to estimate radiation doses from accidents involving 

NRC-exempted products and materials. The methodology considers the following four accident 

exposure scenarios: (1) fires involving the release of radioactive materials from all types of 

exempted products, (2) mechanical resuspension of radioactive materials during cleanup after a 

fire, (3) spills of radioactive materials in liquid or powder form, and (4) crushing of glass tubes 

containing radioactive gases. For these selected accident exposure scenarios, DSRs are 

provided that give the radiation dose per unit quantity of radioactive material at risk in an

A.1-9
accident. Because of the broad range of exemptions, it has been necessary to estimate 

radiation doses on a case-by-case basis when the DSR methodology is not applicable, and 

radiation doses have been estimated on a case-by-case basis in some assessments using 

better available data for a specific accident exposure scenario and product of interest.

A.1-10
Table A.1.1 Factors for Estimating Release of Respirable Size Particles 

During a Spill or Fire 

Type of Factor Value 

Release factor for materials involved in a fire 

Gases 100% 

Solids/powders/liquids 0. 1% 

Solids in protective devices 0.01% 

Release factor for spills of liquids or powders 0.1%

A.1-11
Table A.1.2 Enclosure Volumes and Ventilation Rates

Enclosure Volume Ventilation Rate 

Type of Enclosure (M 3) (volume/h) 

Warehousea 3,000 1 

Storeroom or 300 4 

cargo-handling bay? 

Residencec 450 1 

Bedroomd 27 1 

Watch repair shope 

Small 18 1 

Large 34 1 

Laboratory' 180 6 

Tractor trailerg 87 1 

Large delivery truckh 25 3 

Small delivery truck' 7.1 5 

Automobilej 6.2 5 

a See NUREG/CR-1775 and O'Donnell et al. (1981). Volume corresponds to a warehouse 

measuring approximately 30.5 m x 30.5 m x 3.66 m. 

b See IAEA SS No. 7. Volume corresponds to a storeroom or cargo-handling bay measuring 

10 mx 10 mx 3m. 

c See O'Donnell et al. (1981). Volume corresponds to a residence with a living area of 186 m2 

and a ceiling height of 2.44 meters. 

d See NUREG/CR-0216. Volume corresponds to a room measuring approximately 

3.66 m x 3.05 m x 2.44 m. e See NUREG/CR-0215, NUREG/CR-0216. These reports also provide estimates of the 

enclosure volumes for a jewelry store, catalog store, department store, etc. 

'The volume corresponds to a room with a floor area of 60.4 m2 and a ceiling height of 

3.05 meters. 

g See Etnier and O'Donnell (1979). Cargo area in trailer is assumed to be 13.7 meters long, 2.35 meters wide, and 2.7 meters high. Volume is approximately 87 m3 and ventilation rate is 

likely very low in the trailer. "h See Etnier and O'Donnell (1979). Cargo area of truck is assumed to be 5 meters long, 2.2 meters wide, and 2.3 meters high. Volume is approximately 25 m3 and the ventilation rate 

is assumed to be less than a small delivery truck or automobile. 

'See Etnier and O'Donnell (1979). Cargo area of truck is assumed to be 3.05 meters long, 1.8 meters wide, and 1.3 meters high. Volume is approximately 7.1 m3 and the ventilation rate 

is assumed to be similar to that of an automobile. 

See Etnier and O'Donnell (1979). Passenger area is assumed to be 3.05 meters long, 1.7 meters wide, and 1.2 meters high. Volume is approximately 6.2 m3 and the ventilation rate 

is 5 volume changes per hour.

A.1-12
Table A.1.3 Average Concentrations in Air and Inhalation Intakes of Radioactive 

Material During the 30-Minute Period Following a Fire 

Average 

Ventilation Rate Concentration Individual Intake 

Type of Fire (k) (C) (0 

Transportation firea 4 h-' 1.4xl 0-3 Q m-3 8.6x10 7

- Q 

House fireb 0.25 h- 1 2.1x10- 3 Q m-3 1.3x10-6 Q 

0.5 h-1 2.Ox1 0-3 Q m-3 1.2x10- 6 Q 

1 h-' 1.7x10- 3 Qm-3 1.0x10-6 Q 

2 h-' 1.4x10-3 Q m-3 8.4x10-7 Q 

Warehouse firec 1 h- 1 2.6x10-4 Q m-3 1.6x10-' Q 

2 h-1 2.1x10-4 Qm-3 1.3x10 7

- Q 3 h- 1 1.7x10-4 Q m-3 1.0x10- 7 Q 

4 h-' 1.4x10-4 Q m-3 8.4x10-8 Q 

a Based on an enclosure volume of 300 m3

, a breathing rate of 1.2 m3

/h, and use of a supplied

air respirator with a protection factor of 1,000. b Based on an enclosure volume of 450 m3 , a breathing rate of 1.2 m3

/h, and use of a supplied

air respirator with a protection factor of 1,000. C Based on an enclosure volume of 3,000 mi3, a breathing rate of 1.2 m3

/h, and use of a 

supplied-air respirator with a protection factor of 1,000.

A.1-13
Table A.1 .4 Radiation Dose-to-Source Ratios for a Transportation 

Accident Involving Fire

Dose-to-Source Ratios (DAR) 

Effective Dose Effective Dose Effective Dose 

Equivalent for Equivalent for Equivalent for

Radionuclide Inhalati 

'H (vapor)c 8.3x1011 r 

14C 1.8x 0-12 r 

16SC 2.6x10-" r 

55 Fe 2.3x10'-2 r 

6000 1 .9x10-'0 r' 

63 Ni 2.7x 0-12 r 

"8

Kr (gas) 

'Sr +90Y 1.l1x1l0'r( 

106 Ru + 10 h 4.1 x 10`0 r 

'

09Cd + 109m Ag 9.5x1011 ri 

1

17CS + 1

37

m Ba 2.8x10<'1 r 

1

47 Pm 3.4x10'1 r 

152 Eu 1 .9x10- 10 

204Ti2.1 XloQ12 r 

2

1lOm Bi + 206 TI 6.5x10-10 r( 

210po 8.1 x1 0-9 re 

24'Am 3.8x 0-7 re 

Th (natural)d 1.8x1010 rE 

U (naturaI)e 7.5x10>'1 rE 

U (depleted)' 3.8x10- r 

See following page for footnotes.

on ab Submersion ab Resuspension a

em//pCi 

em/p~Ci 

em/puCi 

em/p~Ci 

em//pCi 

em//.Ci

~M/pLCi 

em/mCi 

em/yCi 

em/pzCi 

em/pCi 

em/pCi 

Dm/pUCi 

~m/gCi 

in/pCi 

n/pUCi 

em/mg 

km/mg 

tm/mgj

2.5x10 '" rem//pCi 

9.3x10>' 0 rem//pCi 

1 .2x1 -9 rem/gCi 

2.4x10 9 rem//Xi 

8.8x 0-12 rem/1gCi 

1.-1 x 101 rem/p~Ci 

4.5x10- 2 rem//pCi 

2.5x1010) rem/gCi 

8.4x 0-14 rem/pCi 

5.3x1010

o rem/puCi 

1 .7x1 Q-l2 rem/puCi 

1.2x10-'0 rem/AiCi 

3.9x10-15 rem/mCi 

7.6x 0-12 rem/yCi 

7.2x101' rem/mg 

6.8x10'5 rem/mg 

6.1x1015 rem/ma

2.0x1011 rem/yCi 

2.9x10-10 rem/puCi 

2.6x10-" rem/gCi 

2.lx 10- rem//pCi 

2.9x101' rem/AtCi 

1 .3x1 Q8 rem//pCi 

4.6x-10-9 rem/gCi 

1.1 x10-9 rem/lzCi 

3.x-1Q10 rem/gCi 

3.7x10-10 rem//pCi 

2.1x-1O-1 rem/puCi 

2.3x10'1 rem/yCi 

7.3x1 Q-8 rem//pCi 

9.Qx1O-8 rem//pCi 

4.2x1 O- rem/gCi 

1.9x10-9 rem/mg 

8.3x10-'0 rem/mg 

4.2x10 10 rem/mg

A.1-14
Footnotes to Table A.1.4

a 1 rem/gCi = 0.27 Sv/Bq; 1 rem/mg = 0.01 Sv/mg. b Values should be reduced by a factor of 10 for fires involving solid radioactive materials 

contained in protective devices (e.g., 241Am in smoke detectors). c Values apply to tritiated water and are increased by a factor of 1.5 to account for absorption 

through the skin (ICRP 30). d Values assume that natural thorium is 20 years old so that decay chain is approximately 87% 

of equilibrium (see Table 3.1.5 of Section 3.1). e Values assume that the mass abundances in natural uranium are 99.2745% by weight 238U, 

0.720% by weight 235U, and 0.0055% by weight 21U (Parrington et al., 1996). 

f Values assume that the mass abundances in depleted uranium are 99.7495% by weight 238

U, 

0.25% by weight 235U, and 0.0005% by weight 2`U (Rich et al., 1988).

A.1-15
Table A.1 .5 Radiation Dose-to-Source Ratios for a Warehouse Fire

Dose-to-Source Ratios (DSR) 

Effective Dose Effective Dose Effective Dose 

Equivalent for Equivalent for Equivalent for

Radionuclide Inhalation',b 

'H (vapor)c 1 .5x10'1 rem/ýiCi 

14C 3.3x1013 rem/MuCi 

46 SC 4.8x 10 Q1 rem/AtCi 

55 Fe 4.3x10'-3 rem/,uCi 

60 Co 3.5x10-" rem/jiCi 

63 Ni 5.0x1013 rem/A~i 

8,5Kr (gas) 

90

Sr + 9Y2.1x10<' 0 rem/pgCi 

1

06 Ru + 10 h7.6x10<' 1 rem/puCi 

19d+ losmAg 1 .8x1 0-' rem/gCi 

'

37Cs + l

37

mBa 5.1 x10-'" rem//pCi 

147Pm 6.2x1012 rem/,uCi 

152 Eu 3.5x1011 rem/MCi 

204 TI 3.8x 0-13 rem/gCi 

2 10m Bi + 206 T1 1 .2x1 O- rem//MCi 

210po 1.5x10-9 rem/MuCi 

241 Am 7.Oxl 0' rem//gCi 

Th (natural)d 3.3x10>'1 rem/mg 

U (natural)' 1 .4x1 0 ' rem/mg 

U (depleted)' 7.1 X1 0- 2 rem/mg 

See following page for footnotes.

Submersion',b Resuspensiona

4.6x1015 rem/MCI 

1.7x10'0 rem/MCi 

2.2x10-10 rem/MXi 

4.4x 101'0 remn/MCi 

1.6x10'-2 rem/MCi 

2.0x10'11 rem/MuCi 

8.4x10-13 rem/MCI 

4.7x10"1 rem/MCI 

1 .5X1 0-14 rem/M.CI 

9.9x10'11 rem/M.Xi 

3.1x10'3 rem//M.Ci 

2.2x10-" rem/M.Ci 

7.2x 0-16 rem/M~i 

1.4x10>-12 rem/MuCi 

1.3x10 -' rem/mg 

1.2x1 0- 15 rem/mg 

1.1x1O 1

5 remn/mg

A.1-16

2.2x 0-12 rem/M-Ci 

3.2x10-" rem/MCi 

2.9x1012 rem/MuCi 

2.3x1010 rem/MzCI 

3.3x 10- " rem/MCI 

1.4x10-9 rem/MCI 

5.0x1010 rem/M.,Ci 

1.2x1010O rem/MuCI 

3.4x10-" rem/MCI 

4.1x101' rem/MCI 

2.3x10-10 rem/MCI 

2.5x 0-12 rem/MuCI 

7.8x1 09 rem/MCi 

9.9x10 9 rem/MCi 

4.6x 0-7 rem/MCI 

2.1x10-10 rem/mg 

9.2x10-11 rem/mg 

4.7x10'1 rem/mai
Footnotes to Table A.1.5

a 1 rem/,uCi = 0.27 Sv/Bq; 1 rem/mg = 0.01 Sv/mg. b Values should be reduced by a factor of 10 for fires involving solid radioactive materials 

contained in protective devices (e.g., 241

Am in smoke detectors). c Values apply to tritiated water and are increased by a factor of 1.5 to account for absorption 

through the skin (ICRP 30). d Values assume that natural thorium is 20 years old so that the decay chain is approximately 

87% of equilibrium (see Table 3.1.5 of Section 3.1). e Values assume that the mass abundances in natural uranium are 99.2745% by weight 2 "U, 

0.720% by weight 235U, and 0.0055% by weight 2

3U (Parrington et al., 1996). 'Values assume that the mass abundances in depleted uranium are 99.7495% by weight 238U, 

0.25% by weight 235U, and 0.0005% by weight 2

3U (Rich et al., 1988).

A.1-17
Table A.1 .6 Radiation Dose-to-Source Ratios for a Residential Fire

Dose-to-Source Ratios (DAR) 

Effective Dose Effective Dose Effective Dose 

Equivalent for Equivalent for Equivalent for 

Inhalation a,b,c Submersion a,b,c Resuspensiona

1.0x10-10 rem//Xi 

2.1 x1 V12 rem/,uCi 

3.0x`10-11 rem/M-Ci 

2.7x10'-2 rem/MCi 

2.2x10-'0 rem/M.Ci 

3.1x1lO-2 rem/M2Ci

Radionuclide 

'H (vapor)d 

14C 

46 SC 

55 Fe 

600Co 

63 Ni 

`5Kr (gas) 

90S r + 0 

1

0 6 Ru+ 1

06 Rh 

'

09Cd + losmAg 

13705 + 137 m Ba 

204 TI 

2 10m Bi + 206 TI 

210PO 

241

Am 

Th (natura!)e 

U (natural)' 

U (depleted)9

3.1lX1lO 14 rem//MCi 

1.2x10-9 rem/pyCi 

1 .5x1 Q-9 rem//MCi 

3.Qxl O- rem/,uCi 

1.1x10'11 rem/MuCi 

1.3x1 0-" rem/jiCi 

5.8x10'-2 rem4tCi 

3.3x 10-10 rem/puCi 

1.0x1013 rem//-cii 

6.5x10-10 rem/ktCi 

2.1 X1O -12 rem/4uCi 

1.5x10'10 rem/puCi 

5.0x10-15 rem/MCi 

9.8x10'-2 rem/gCi 

1.5x1013 rem/mg 

8.3x10 15 rem/mg 

7.4xl0'5 rem/mcj

1.1x10'1 rem/M~i 

1.6x1010o rem4pCi 

1.4x1QO<" rem/yCi 

1.1x10-9 rem/jLCi 

1.6x10<' 1 rem/MuCi 

6.7xl109 rem/MuCi 

2.5x 10-9 rem/,aCi 

5.7x10<' 0 rem/MuCi 

1.7x10'0 rem//pCi 

2.Ox1QO<' 0 rem/p~Ci 

1 .2x1 O- rem//MCi 

1.2x10<' 1 rem/gCi 

3.9x1 0- remn/MZi 

4.9x 0-8 remn/MCi 

2.3x1 0-6 rem/MuCi 

1.1x10-9 rem/mg 

4.5x10 10 rem/mg 

2.3x10-10 rem/ma

See following page for footnotes.

A.1-18

1 .3x-10-9 rern/jzCi 

4.8x10>'0 rem/puCi 

1.1x10'0O rem/pzCi 

3.2x10'1 rem/,uCi 

3.9x10 11 rem/pzCi 

2.2x1010 rem/M-Ci 

2.4x1Q012 rem/guCi 

8.OxlT9' rem/gCi 

9.4x 10O' rem/MuCi 

4.4x1 0-7 rem/gCi 

2.2x10-10 rem/mg 

8.7x1011 rem/mg 

4.4x10-11 rem/mg
Footnotes to Table A.1.6

a 1 rem//uCi = 0.27 Sv/Bq; 1 rem/mg = 0.01 Sv/mg. b Values should be reduced by a factor of 10 for fires involving solid radioactive materials 

contained in protective devices (e.g., 241

Am in smoke detectors). C Values apply to a firefighter. To estimate radiation doses to a person escaping from a fire or a 

neighborhood hero trying to rescue a person from a fire, multiply DSRs for inhalation by 410 

and DSRs for submersion by 0.39 (see Section A.1.7.2). d Values apply to tritiated water and are increased by a factor of 1.5 to account for absorption 

through the skin (ICRP 30). e Values assume that natural thorium is 20 years old so that decay chain is approximately 87% 

of equilibrium (see Table 3.1.5 of Section 3.1). 'Values assume that the mass abundances in natural uranium are 99.2745% by weight =LU, 

0.720% by weight 235U, and 0.0055% by weight 2 4U (Parrington et al., 1996). 

g Values assume that the mass abundances in depleted uranium are 99.7495% by weight 2-8U, 

0.25% by weight 235

U, and 0.0005% by weight 2 4U (Rich et al., 1988).

A.1-19
Table A.1.7 Average Concentrations in Air and Individual Intakes From Inhalation 

During Various Exposure Times Following a Spill in a Laboratory-Type Room a 

Ventilation Rate Exposure Time Average Concentration Individual Intake (A) (0) (C) M• 

6 h-1 15 min 2.9x10-3 Qm-3 8.6x10 4 Q 

30 min 1.8x10- 3 Q m- 3 1.0x10-3 Q 

1 h 9.2x10-4 Qm-3 1.1x10-3 Q 

9 h-1 15 min 2.2x10-3 Qm-3 6.6x10-4 Q 

30 min 1.2x10-3 Q m-3 7.3x10 4 Q 

1 h 6.2x10- 3 Q m-3 7.4x10-3 Q 

12 h-1 15 min 1.8x10-3 Qm-3 5.3x10 4 Q 

30 min 9.2x10- 4 Q M-3 5.5x10-4 Q 

1 h 4.6xl 0-4 Q m-3 5.5x1 0-4 Q 

a Assumes an enclosure volume of 180 m3 and a breathing rate of 1.2 m3

/h without a respirator.

A. 1-20
Table A.1.8 Radiation Dose-to-Source Ratios for a Spill of a Liquid 

or Powder in a Laboratory-Type Room

Dose-to-Source Ratios (DSRs) 

Effective Dose Equivalent Effective Dose Equivalent 

Radionuclide for Inhalationa for Ingestiona 

3

H (liquid) 9.6x10-11 rem/yCib 6.4x10-6 rem/MCiF 

14C 2.1x10-9 rem/,Ci 2.1 xl 0-7 rem/gCi 

46SC 3.0xl 0-8 rem//MCi 6.4xl 0-' rem/yCi 

5 5Fe 2.7x 10-9 rem//,tCi 6.1 x 10 8 rem/MCi 

60C0 2.2x10-7 rem/,Ci 1.0x10-6 rem/pCi 

63Ni 3.1 x10-9 rem/,uCi 5.8x 10 8 rem/,uCi 

9OSr + 9oy 1.3xl 0- 6 rem/MCi 1.4xl0- rem/pCi 

°

06Ru + 106

Rh 4.8xl 0-7 rem/pCi 2.7xl 0-6 rem//, Ci 

1°90d + 109 mAg 1.1X10- 7 rem/gCi 1.3xl 0- 6 rem/pCi 

137Cs + 137mBa 3.2x10-8 rem//,Ci 5.0xl 0-6 rem/puCi 

147pm 3.9xl 0-8 rem/MCi 1.1x10- 7 rem/pCi 

15 2Eu 2.2xl 0' rem/gCi 6.5xl 0-7 rem/pCi 

204

TI 2.4xl 0- 9 rem//,Ci 3.4xl 0-7 rem/pCi 

2lomBi + 206

TI 7.6xl 0-6 rem/,Ci 9.6xl 0-6 rem/pCi 

21°Po 9.4xl 0-6 rem//,tCi 1.9x1 0-4 rem//pCi 

24 1

Am 4.4xl 0-4 rem/pCi 3.6xl 0-4 rem/pCi 

Th (natural)d 2.1xl 0-7 rem/mg 5.1 xl 08 rem/mg 

U (natural)e 8.7xl 0-8 rem/mg 1.9x 10 8 rem/mg 

U (depleted)' 4.4x10-8 rem/m-q 1.0xlO-8 rem/ma

See following page for footnotes.

A. 1-21
Footnotes to Table A.1.8

a 1 rem//.zCi = 0.27 Sv/Bq; 1 rem/mg = 0.01 Sv/mg. b Value applies to tritiated water and is increased by a factor of 1.5 to account for absorption 

through the skin (ICRP 30). c Value applies to intake of tritiated water through skin rather than direct ingestion of tritiated 

water spilled on the skin (see Section A.1.6). d Values assume that natural thorium is 20 years old so that decay chain is approximately 87% 

of equilibrium (see Table 3.1.5 of Section 3.1). e Values assume that mass abundances in natural uranium are 99.2745% by weight 238U, 

0.720% by weight 235U, and 0.0055% by weight 2

34U (Parrington et al., 1996). f Values assume that mass abundances in depleted uranium are 99.7495% by weight 238U, 

0.25% by weight 235U, and 0.0005% by weight 2

'U (Rich et al., 1988).

A. 1 -22
Table A.1.9 Radiation Dose-to-Source Ratios for Crushing of Glass Tubes 

Dose-to-Source Ratios (DSRs) 

(rem/4Ci)a 

Enclosureb Exposure Time 3Hc 1Kr(d 220Rne

Warehouse 

Storeroom or 

cargo-handling bay 

Residence 

Bedroom 

Large watch repair 

shop 

Small watch repair 

shop 

Laboratory

30 min 

lh 

2h 

4h 

8h 

15 min 

30 min 

lh 

2h 

30 min 

lh 

2h 

4h 

8h 

30 min 

lh 

2h 

4h 

8h 

30 min 

lh 

2h 

4h 

8h 

30 min 

lh 

2h 

4h 

8h 

15 min 

30 min 

lh

1.5x10 8 

2.4x1 0-8 

3.3x1 0-8 

3.8x10-8 

3.8x10-8 

6.1x10-8 

8.3x1 0-8 

9.4x1 08 

9.6x10-8 

1.0xl 0-7 

1.6x10-7 

2.2x1 0-7 

2.5x 10-7 

2.6x1 0-7 

1.7x10-6 

2.7x10-6 

3.7x1 0-6 

4.2x 106 

4.3x1 0-6 

1.3x1 0-6 

2.1x10-6 

2.9x1 0-6 

3.3x1 0-6 

3.4x 106 

2.5x1 0-6 

4.Oxl 0-6 

5.5x10-6 

6.3xl 0-6 

6.4x10-6 

8.3x10-8 

1.Oxl 0-7 

1.1x10-7

See following page for footnotes.

A.1-23

4.4x10 10 

7.1x10-10 

9.7x 10`0 

1.1x10-9 

1.1x10-9 

1.8xl 0-9 

2.4x10-9 

2.7x10-9 

2.8xl 0-9 

2.9x1 0-9 

4.7xl 0-9 

6.4x10-9 

7.3x1 0-9 

7.5x10-9 

4.9xl 0-8 

7.8x10-8 

1.1x10-7 

1.2x1 0-7 

1.2x10-7 

3.9x10-8 

6.2x10-8 

8.5x10-8 

9.7x1 0-8 

9.9x10-8 

7.3x1 0-8 

1.2x10-7 

1.6x10-7 

1.8x10-7 

1.9x10-7 

2.4x1O-9 

2.9x10-9 

3.1x10-9

3.4x10-8 

5.4x10 8 

7.3x10-8 

8.2x10-8 

8.3x1 0-8 

1.4x10-7 

1.9x10-7 

2.1x10-7 

2.2x1 0-7 

2.3x1 0-7 

3.6xl 0-7 

4.9x10-7 

5.4x10-7 

5.5x1 0-7 

3.8x10'6 

6.Oxl 0-6 

8.1x10 6 

9.1x10-6 

9.2x10-6 

3.Ox1 06 

4.8x10-6 

6.5x10-6 

7.2x10-6 

7.3x10-6 

5.7x10-6 

9.1x10-

6 

1.2x10 

1.4x10 5

1.4x10 5

1.9x10-7 

2.3x10-7 

2.4x10-7
Footnotes to Table A.1.9

a 1 rem//.LCi = 0.27 Sv/Bq. 

b See Table A.1.2 for enclosure volumes and ventilation rates. 

c Values for internal dose from inhalation and absorption of tritiated water vapor (HTO) through 

the skin are based on an effective dose equivalent (EDE) of 9.6xl 0-5 rem/UCi of HTO in air 

(see Table 2.1.2 of Section 2.1, and EPA-520/1-88-020). 

d Values for external dose from submersion in 85

Kr are based on an EDE rate of 0.14 mSv/yr 

(1.4x1 0-2 rem/yr) plus 1% of the skin dose-equivalent rate of 1.5 rem/yr per MCi/m 3 of 85Kr in air 

(see Table 2.1.2 of Section 2.1, and EPA-402-R-93-081). 

e Values for internal dose from inhalation of 22

°Rn and its progeny are based on an (1) EDE rate 

of 1.6xl 03 rem/yr per MCi/m 3 of 2°Rn in radioactive equilibrium with its progeny in indoor air and 

(2) equilibrium factors calculated as ratios of time-averaged concentrations of 212Pb to 2°Rn in 

indoor air of various structures (see Section 3.1 and ICRP 50).

A. 1-24
A.2 GENERIC DISPOSAL METHODOLOGY 

A.2.1 Introduction 

This appendix presents a generic methodology for estimating radiation doses from disposal of 

exempted amounts of source or byproduct materials. Doses are estimated for the following three disposal options: municipal landfills, municipal waste incinerators, and recycling in metals. 

For each disposal option, groups of exposed individuals and populations are defined, including 

workers associated with operations at landfills, incinerators, and metal smelters and members 

of the public who could be exposed in a variety of ways, depending upon the particular option. 

For disposal in landfills and incinerators, both individual and collective doses are calculated. 

However, because recycling of metals should be an unusual occurrence for most exempted 

materials, only individual doses are estimated for this option. 

For each group of exposed individuals and populations for an assumed disposal option, 

assumed exposure pathways, which generally include external exposure, inhalation, and 

ingestion, are defined. Then, for each exposure pathway, radionuclide-specific dose-to-source 

ratios (DSRs) are calculated for the exposed individuals and populations. The DSRs give 

effective dose equivalents (EDEs) per unit activity of radionuclides disposed for the assumed 

disposal option and exposure pathway. Doses then are estimated from the relationship 

Hij =DSRij x A j,(1 

where H,1 = EDE from exposure to radionuclide ifor exposure pathwayj, 

DSR = dose-to-source ratio for the particular radionuclide and exposure pathway, and 

A, = assumed activity of the particular radionuclide disposed for the assumed 

option. 

The DSRs are the quantities calculated in the generic disposal methodology. The DSRs then 

are applied to assumed activities of radionuclides to estimate doses to individuals and 

populations. 

The following section describes the three disposal options assumed in developing the generic 

disposal methodology, including the groups of exposed individuals and populations for each 

option and the exposure pathways assumed for each group. The next three sections present 

the models and parameter values for calculating the DSRs for each of the three disposal 

options and the calculated DSRs for each radionuclide and exposure pathway. The last section 

illustrates the application of the calculated DSRs to the estimation of individual and collective 

doses from disposal of exempted amounts of radionuclides. 

A.2.2 Description of Disposal Options 

This section describes the exposure scenarios and associated exposure pathways for landfill 

disposal, incineration, and metal recycling assumed in the generic disposal methodology for the 

purpose of estimating doses to exposed individuals and populations.

A.2-1
A.2.2.1 Disposal in Municipal Landfills

More than half of all municipal solid waste generated in the United States is sent to landfills for 

disposal (Environmental Protection Agency (EPA) 530-R-9-042). Therefore, except in unusual 

cases in which particular items containing exempted amounts of radionuclides are not expected 

to enter municipal waste streams, disposal in municipal landfills should be a common 

occurrence. 

For disposal of exempted amounts of radioactive materials in municipal landfills, the following four groups of individuals are assumed to be exposed: (1) waste collectors, (2) workers at the 

landfills, (3) off-site members of the public residing near the landfills, and (4) future on-site 

residents at the landfills. The assumed exposure pathways for these groups are described in 

the following paragraphs. 

A.2.2.1.1 Waste Collectors 

Waste collectors are individuals who collect waste from the generating site, haul the waste to 

garbage trucks, and transport the waste to landfills. Exposure to waste collectors are assumed 

to occur primarily during hauling of waste to garbage trucks in small containers. Exposure to 

waste collectors during transport to landfills should be considerably less than during waste 

collection because (1) the exposure time during transport should be much less than during collection, (2) the distance between a waste collector and the sources should be greater during 

transport than while hauling waste containers, and (3) garbage trucks should provide greater shielding from external exposure than waste containers. Waste collectors are assumed to 

receive exposures from the following three pathways: (1) external exposure to radionuclides in 

the waste containers, (2) inhalation of radionuclides emitted from the waste containers into the 

air, and (3) ingestion of radionuclides in the waste. 

A.2.2.1.2 Landfill Workers 

Workers at landfills are individuals who are located on top of the waste pile during operations and who perform tasks such as dumping of waste, grading of waste following dumping, and 

covering of the waste at periodic intervals. Exposure to other workers at a landfill should be 

considerably less than exposure to workers on the waste pile, primarily because other workers 

would be located at much larger distances from any sources. Workers at landfills are assumed 

to receive exposures from the following three pathways: (1) external exposure to radionuclides 

in the waste pile, (2) inhalation of radionuclides suspended from the waste pile into the air, and 

(3) ingestion of radionuclides in the waste pile. 

A.2.2.1.3 Off-Site Members of the Public 

For off-site members of the public who reside near landfills, two different exposure scenarios 

are considered. The first scenario, which would occur only during landfill operations, involves 

releases of radionuclides into the air and subsequent atmospheric transport to off-site locations. 

For atmospheric releases during landfill operations, off-site residents are assumed to receive 

exposures from the following four pathways: (1) inhalation of airborne radionuclides, (2) external exposure to airborne radionuclides, (3) external exposure to radionuclides deposited 

on the ground surface, and (4) ingestion of food products contaminated by deposition onto the 

ground surface.

A.2-2
The second exposure scenario for off-site members of the public who reside near landfills, 

which is assumed to occur only after landfills are closed, involves releases of radionuclides into 

groundwater and subsequent transport to a nearby municipal well. Exposures for this scenario 

would occur at times considerably later than the exposures from atmospheric releases during 

landfill operations described above. Therefore, the exposed individuals and populations in the 

two scenarios would not be the same. For releases to groundwater, off-site residents are 

assumed to receive exposures from the pathway that involves ingestion of radionuclides in 

drinking water obtained from a well. Based on a previous generic assessment for water 

releases (Cook and Hunt, 1994), other potential exposure pathways are assumed to be 

insignificant. 

A.2.2.1.4 Future On-Site Residents 

At some time after closure of a landfill, members of the public are assumed to establish 

permanent residency on the landfill site. A suburban housing development is assumed, in 

which no on-site sources of drinking water are established. Exposure to on-site residents are 

assumed to result from the uncovering of waste during excavation at the site. Residents are 

assumed to receive exposures from the following three pathways: (1) external exposure to 

radionuclides in the waste during indoor and outdoor residence on the site, (2) inhalation of 

radionuclides suspended from the waste into the air during indoor and outdoor residence on the 

site, and (3) ingestion of radionuclides in the waste. Based on the assumption that a suburban 

housing development is established at the landfill site, no other exposure pathways are 

assumed to occur. 

A.2.2.2 Disposal in Municipal Incinerators 

A substantial fraction of all municipal solid waste generated in the United States is sent to 

incinerators for disposal (EPA-530-R-94-042). Therefore, except in unusual cases in which 

particular items containing exempted amounts of radionuclides are not expected to enter 

municipal waste streams, disposal by incineration should be a common occurrence. 

For disposal of exempted amounts of radioactive materials by incineration, the following three 

groups of individuals are assumed to be exposed: (1) waste collectors, (2) workers at the 

incinerators, and (3) off-site members of the public residing near the incinerators. The 

assumed exposure pathways for these groups are described in the following paragraphs. 

A.2.2.2.1 Waste Collectors 

For waste collectors at incinerators, the assumed exposure scenario and exposure pathways 

are the same as those described for waste collectors at landfills in Appendix A.2.2.1.1. 

Therefore, waste collectors are assumed to receive exposures from the following three 

pathways: (1) external exposure to radionuclides in the waste containers, (2) inhalation of 

radionuclides emitted from the waste containers into the air, and (3) ingestion of radionuclides 

in the waste. 

A.2.2.2.2 Incinerator Workers 

Workers at incinerators are individuals who engage in sweeping or other cleanup activities 

while located at the edge of a partially enclosed tipping area where garbage trucks unload

A.2-3
waste at the facility. Exposure to workers during waste unloading in the tipping area are 

assumed to be substantially higher than exposure to workers during other operations at the 

incinerators. Workers at incinerators are assumed to receive exposures from the following three pathways: (1) external exposure to radionuclides in the waste pit, (2) inhalation of 

radionuclides suspended from the waste pit into the air, and (3) ingestion of radionuclides in the 

waste. 

A.2.2.2.3 Off-Site Members of the Public 

Off-site members of the public who reside near incinerators are assumed to receive exposures 

from stack releases of radionuclides into the air following waste incineration and subsequent atmospheric transport to off-site locations. The assumed exposure scenario and exposure pathways for airborne releases from an incinerator are the same as those described for 

airborne releases during landfill operations in Appendix A.2.2.1.3. Therefore, off-site residents 

are assumed to receive exposures from the following four pathways: (1) inhalation of airborne 

radionuclides, (2) external exposure to airborne radionuclides, (3) external exposure to 

radionuclides deposited on the ground surface, and (4) ingestion of food products contaminated 

by deposition onto the ground surface. 

Off-site releases of airborne radionuclides from incineration facilities also could occur during unloading of waste from garbage trucks into the tipping area. These releases would result in 

the exposure pathways for off-site residents listed above. However, because of the general 

concern about airborne releases of hazardous substances at waste incinerators, the partially enclosed tipping area normally is ventilated so that a negative pressure, compared with the 

outdoor air pressure, is maintained (Phone call, S. J. Levy, Office of Solid Waste, U.S. 

Environmental Protection Agency, Washington, DC, February 1997). This precludes substantial releases into the atmosphere during dumping operations and any other activities in 

the tipping area. Therefore, releases during waste dumping should be unimportant compared 

with stack releases during incineration. 

Following incineration of waste, ash is removed for final disposal. However, doses from 

disposal of incinerator ash are not considered in this assessment, primarily because incinerator 

ash normally must be managed separately from municipal waste and would not be sent to 

municipal landfills (EPA-530-R-94-042). Incinerator ash normally is managed as hazardous 

waste due, for example, to the presumed presence of toxic heavy metals, and disposal in a 

permitted facility for hazardous waste is required. Because of the stringent requirements for 

treatment and disposal of hazardous waste under the Resource Conservation and Recovery Act (RCRA) specified in 40 CFR 264, disposal of incinerator ash as hazardous waste should 

result in doses substantially less than the doses from disposal of nonhazardous waste in 

municipal landfills. 

A.2.2.3 Metal Recycling 

A substantial fraction of all municipal solid waste generated in the United States is recovered for 

recycling (EPA-530-R-94-042). However, most of the recovered and recycled materials 

include items such as paper and paper products, plastic, glassware, and aluminum and other 

metal containers that would not contain radioactive materials, and recycling of most items 

containing exempted amounts of radioactive materials is not expected to be a common 

occurrence.

A.2-4
In this assessment, recovery and recycling of items containing exempted amounts of 

radioactive materials are assumed to occur only for the purpose of recovering ferrous metals 

and alloys (e.g., steel). This assessment is not concerned with recycling of exempted items 

when the intent is to recover and reuse the radioactive material itself, because (1) this activity is 

not a form of disposal, (2) it normally would not result in the introduction of radioactive material 

into some other product as incidental contamination, and (3) it would be carried out by licensees of the Nuclear Regulatory Commission (NRC) or an Agreement State. 

For exempted items that could be assumed to be recycled for the purpose of recovering ferrous 

metals, the items are assumed to be sent to a metal (i.e., steel) smelter. The following three 

groups of individuals are assumed to be exposed: (1) workers at smelters, (2) off-site members 

of the public residing near smelters, and (3) members of the public who use recycled products containing radioactive material. Assumed exposure pathways for these groups are described in 

the following paragraphs. 

A.2.2.3.1 Smelter Workers 

Based on a previous assessment (Hill et al., 1995), the individuals at metal smelters who are 

assumed to receive the highest doses are slag workers. These workers are assumed to 

receive exposures from the following three pathways: (1) external exposure to radionuclides in 

slag, (2) inhalation of radionuclides emitted from slag into the air, and (3) ingestion of 

radionuclides in slag. 

A.2.2.3.2 Off-Site Members of the Public 

Off-site members of the public who reside near smelters are assumed to receive exposures from stack releases of radionuclides into the air following smelting and subsequent atmospheric 

transport to off-site locations. The assumed exposure scenario and exposure pathways for 

airborne releases from a smelter are the same as those described in Appendix A.2.2.1.3 for 

airborne releases during landfill operations. Therefore, off-site residents are assumed to 

receive exposures from the following four pathways: (1) inhalation of airborne radionuclides, 

(2) external exposure to airborne radionuclides, (3) external exposure to radionuclides 

deposited on the ground surface, and (4) ingestion of food products contaminated by deposition 

onto the ground surface. 

A.2.2.3.3 Users of Recycled Products 

Members of the public are assumed to receive exposures during use of products containing recycled metal. During use of contaminated products, members of the public are assumed to 

receive exposures from the pathway that involves external exposure to radionuclides in the 

product. Inhalation and ingestion of radionuclides in recycled metal products would not 

normally occur (Hill et al., 1995) and, thus, is not considered in this assessment. 

A.2.3 Dose Assessment for Disposal in Landfills 

This section presents the models and parameter values for estimating individual and collective 

doses from disposal of radioactive material in municipal landfills, and the results of the dose 

assessment in the form of doses per unit activity of radionuclides disposed in all landfills

A.2-5
(i.e., the DSRs in Equation (1) are tabulated). The groups of individuals considered in the dose 

assessment and their associated exposure pathways are described in Appendix A.2.2.1. 

A.2.3.1 Waste Collectors 

Waste collectors at municipal landfills are assumed to receive external, inhalation, and 

ingestion exposures while hauling waste in containers from collection sites to a garbage truck. 

The dose assessment for waste collectors is described in the following paragraphs. 

A.2.3.1.1 External Exposure to Individuals 

The EDE to individual waste collectors from external exposure is calculated by assuming that a 

unit activity of each photon-emitting radionuclide is uniformly distributed in a waste container, 

which is assumed to be a cylinder with a height of 0.9 meter and a radius of 0.38 meter. The 

uncompacted waste is assumed to have an average density over the container volume of 

0.4 g/cm3, and the self-shielding provided by the waste is taken into account by assuming that 

the waste resembles water in its shielding properties. The shielding provided by the walls of the 

container is taken into account by assuming that the wall thickness is 0.32 cm and that the wall 

material, which normally is plastic, also resembles water in its shielding properties. Based on 

these assumptions, the EDE rate near a waste container for a unit activity of 1 microcurie (UCi) 

(37 kilobecquerel (kBq)) of each radionuclide in the waste was calculated using MicroShield 

(Computer Codes, Grove Engineering, 1996). 

In estimating the annual EDE to an individual waste collector, the following exposure conditions 

were assumed. First, the collector was assumed to be located at a distance of 0.3 meter from 

the surface of the waste container. Second, exposure to a single waste container was 

assumed to occur for 4 hours, based on information that waste collectors typically work this 

long before unloading the contents of a garbage truck (Phone call, J. Bailey, Browning-Ferris 

Industries, Knoxville, TN, July 1994). Thus, the annual individual dose was calculated by 

assuming that all of the annual disposals of radionuclides in landfills occur during a single waste 

collection. Finally, in calculating the dose to a waste collector at a single landfill, the annual 

disposal of a unit activity of each radionuclide in all landfills was assumed to be distributed 

equally among 3500 operating landfills (EPA-530-R-96-006). 

A.2.3.1.2 Inhalation Exposure to Individuals 

The annual EDE to an individual waste collector from inhalation exposure is estimated by 

assuming that the radioactive material is in a readily dispersible form and could be released into 

the air during waste collection. The annual individual dose, Hi, from an annual disposal of a 

unit activity, A,, of 1 kiCi (37 kBq) of each radionuclide in all landfills is given by 

Hi/A, (rem/ = =Ci) (1/NL) x (1/M,) x La x fax Uax TxT D , (2) 

where NL = number of operating landfills, 

Mc = mass of waste per waste container (g), 

La = atmospheric mass loading of waste emitted from waste container (g/m3

), 

fa = respirable fraction of airborne material,

A.2-6
Ua = breathing rate for waste collector (m3

/h), 

T = exposure time for waste collector (h), and 

Dý, h,= inhalation dose coefficient for radionuclide i (rem/4UCi). 

As in the analysis for external exposure described in the previous section, the number of 

operating landfills, N, Iin which disposals of radionuclides are assumed to be distributed 

equally is 3500. The annual inhalation dose to an individual waste collector can be calculated 

by assuming that all exposures occur during a single collection trip containing 1 year's disposals of a radionuclide. The assumed values of the other parameters in Equation (2) are described 

as follows: 

0 The mass of waste per container, MC, was 1.4x1 05 g, based on the assumed waste 

density of 0.4 g/cm3 and the dimensions of a waste container given in 

Appendix A.2.3.1.1. 

* The atmospheric mass loading of waste, L a, was 4x10-5 g/m3, as described below. 

* The respirable fraction of the airborne material, fa, was 0.7 (EPA, RAE-9232/1-2). 

* The breathing rate, Ua, was 1.2 m3/h, which is a value appropriate for light activity (see 

Appendix A.1). 

0 The exposure time for a single collection trip, T, was 4 hours (see Appendix A.2.3.1.1). 

0 The inhalation dose coefficient (Dih) for each radionuclide was obtained from current 

Federal guidance (EPA-520/1-88-020). 

0 For tritium (3H) the inhalation dose factor has been increased by a factor of 1.5 to 

account for absorption through the skin. 

No data are available for estimating the atmospheric mass loading of waste emitted from a 

container during waste collection. The value assumed in this assessment corresponds to an 

average background dust loading (Anspaugh et al., 1975). The release of readily dispersible material to the atmosphere during waste collection could be greater than the average release of 

naturally occurring material on the ground surface. However, much of the waste could be 

contained, for example, in plastic bags, and releases from a small source should be dispersed 

away from the waste collector by prevailing winds for some fraction of the time, even when the 

collector is close to the container. 

A.2.3.1.3 Ingestion Exposure to Individuals 

The annual EDE to an individual waste collector from ingestion exposure is estimated by assuming that the radioactive material is in a readily dispersible form and could be transferred 

to the hands of the individual during waste collection. The annual individual dose, H,, from an 

annual disposal of a unit activity, A,, of 1 /aCi (37 kBq) of each radionuclide i in all landfills is 

given by 

H,/A i (rem/yCi) = (1/NL) X (1/Me) X Uing x T x Dingi, (3)

A.2-7
where the factor 1/NL again represents the assumption that the annual disposals of 

radionuclides are distributed equally among all operating landfills, the factors M, and Tare the 

mass of waste per waste container and exposure time for the waste collector described with 

Equation (2) in the previous section, and the other factors are described as follows: 

Uing = ingestion rate of waste for waste collector (g/h), and 

Ding, i = ingestion dose coefficient for radionuclide i (rem/,"Ci). 

As in the analyses for external and inhalation exposure described previously, the annual 

ingestion dose to a waste collector can be calculated by assuming that all exposures occur 

during a single collection trip containing 1 year's disposals of a radionuclide. The assumed 

values of the parameters in Equation (3) that were not presented with Equation (2) are the 

ingestion rate of waste, Ujg (6x10-3 g/h), which is a value appropriate for commercial or 

industrial activities (EPA, OSWER Directive 9285.6-03), and the ingestion dose coefficient 

(Ding) for each radionuclide which was obtained from current Federal guidance 

(EPA-520/1-88-020). 

A.2.3.1.4 Collective Dose for Waste Collectors 

The calculations of annual individual dose described in the previous three sections assume that 

only one waste collector at each landfill collects waste containing radioactive material. This 

assumption would overestimate the dose to an average waste collector when more than one 

collector is involved at a landfill and the radioactive materials are distributed randomly in all 

waste collections. However, given that a typical (i.e., median) landfill receives about 2.5x1 06 kg 

of waste per year (EPA/530-SW88-034) and that a normal garbage truck with a capacity of 

about 20 m3 (Phone call, J. Bailey, Browning-Ferris Industries, Knoxville, TN, July 1994) can 

haul about 2x1 04 kg of waste per shipment, assuming a density of compacted waste of about 

1 g/cm 3, a typical landfill evidently could be serviced by very few trucks and, thus, very few 

waste collectors. Therefore, it is not unreasonably pessimistic to assume only a single exposed 

waste collector per typical landfill, even though this would not be the case at unusually large landfills. Based on the assumption that only a single waste collector is exposed per landfill, the 

annual collective dose to waste collectors from 1 year's disposals of a radionuclide in all landfills 

is obtained by multiplying the annual individual dose by the total number of operating landfills, 

which again is assumed to be 3500 (EPA-530-R-96-006). 

A.2.3.1.5 Results of Dose Calculations 

The annual individual and collective EDEs to waste collectors at municipal landfills from 1 year's 

disposals of a unit quantity of 1 pCi (37 kBq) of each radionuclide in all landfills estimated in this 

assessment (i.e., the DSRs in Equation (1)) are presented in Tables A.2.1 and A.2.2. In 

applying the DSRs to disposal of particular items containing radioactive material, the following 

points should be noted. 

First, if a particular item could be sent to either landfills or incinerators for disposal, which is 

often assumed to be the case, the DSRs should be reduced by the fraction of the total annual 

disposals sent to landfills. Recent data indicate that the amount of waste sent to landfills is 

about four times the amount sent to incinerators (EPA-530-R-94-042). Therefore, if the input 

to the dose assessment is an assumed total activity of a radionuclide disposed per year in all

A.2-8
landfills and incinerators, the DSRs for all exposure pathways should be reduced by a factor of 

0.8 to take into account the fraction of the disposed activity sent to landfills. 

Second, the DSRs for inhalation and ingestion assume that the radioactive materials are in a 

readily dispersible form, but this would not be the case for many items. Therefore, depending 

on the physical form of the particular items of concern, the assessor could reduce the DSRs for 

inhalation and ingestion whenever the radioactive materials should be significantly less 

dispersible than loose materials in trash. The following guidelines for reducing the DSRs for 

inhalation and ingestion in these cases are suggested: 

For some exempt items, distribution and use may be limited, and it may not be 

conservative to assume that the annual disposal occurs uniformly over all 3500 landfills. 

For this situation, an adjustment should be applied to reflect localized use and disposal. As a simple adjustment, if the assumed number of items to be disposed of annually is 

less than 3500, which is the assumed number of disposal sites, the DSRs should be 

increased by the ratio of 3500 divided by the number of items. 

For radioactive materials that should be considerably less dispersible than loose 

materials in trash, but nonetheless could be dispersed to some extent, the DSRs for 

inhalation and ingestion may be reduced by a factor of 10. Examples of this case might include radioactive materials in the form of plated foils or solid items that are easily 

breakable or crushable into small pieces. 

For items that should be nondispersible during normal waste collection activities, the 

DSRs for inhalation and ingestion may be assumed to be zero (0). Examples of this 

case might include large solid metal forms or radioactive materials dispersed in 

substantial glass or ceramic forms. 

If no correction for dispersibility is applied, then, for any radionuclide, the DSR for either 

individual or collective dose from all exposure pathways is the sum of the DSRs for external 

exposure, inhalation, and ingestion. 

A.2.3.2 Workers at Landfill 

Workers at municipal landfills are assumed to receive external, inhalation, and ingestion 

exposures while located on top of a waste pile. The dose assessment for landfill workers is 

described in the following paragraphs. 

A.2.3.2.1 External Exposure to Individuals 

The annual EDE to an individual landfill worker from external exposure is estimated by assuming that the source is an infinitely thick, uniformly contaminated volume of soil-equivalent material and that the worker is operating heavy equipment on top of exposed waste at the 

working face of the landfill. The annual individual dose, Hi, from an annual disposal of a unit 

activity, A i, of 1 gaCi (37 kBq) of each photon-emitting radionuclide i in all landfills is given by 

H iA i (rem//-Ci) = (1/N L) x (11MJ) x fe., x f h x D ext, i,7 (4)

A.2-9
where the factor 1/NL represents the assumption described in Appendix A.2.3.1 .1 that the 

annual disposals of radionuclides are distributed equally among all operating landfills, and the 

other factors are described as follows: 

M, = mass of waste disposed in landfill annually (g/yr), 

fex = fraction of the year during which exposure occurs, 

fst, = shielding factor while operating heavy equipment, and 

De•. i = external dose coefficient for radionuclide i (rem/yr per /Ci/g). 

The assumed values of the parameters in Equation (4), except for NL, which again is assumed 

to be 3500, are described as follows: 

0 The mass of waste disposed in a landfill annually, M,, was 2.5x10 9 g, as described 

below. 

0 The fraction of the year during which exposure occurs, fex, was 0.18, based on an 

assumed time spent working on the waste pile of 1600 h/yr. 

0 The shielding factor while operating heavy equipment, f 5h, was 0.75. 

a The external dose coefficient (D, 4 ) for each radionuclide for an infinitely thick volume 

source in soil was obtained from current Federal guidance (EPA 402-R-93-081). 

The mass of waste disposed in a landfill was assumed to be the median value for all landfills 

(EPA-520/1-88-020). The median value was used, rather than the average, because the 

distribution of disposals in landfills is highly skewed and about 84% of all landfills receive less 

than the average amount of waste (EPA-520/1-88-020). Use of the median value also gives 

higher estimates of dose. 

For 8

"Kr, external exposure was estimated by assuming that half of the activity would be 

released into the air during landfill operations, due to breakage of half of the containers for this 

radionuclide, and half would be retained in the waste in intact containers. Therefore, the 

external dose from 85

Kr retained in the waste would be half of the value calculated as described 

above. The 85

Kr released into the air also would result in external exposure to workers from 

submersion in an atmospheric cloud. However, if the release occurs over the working face of 

the landfill with an assumed area of 2100 m2 (i.e., an area of about 46 m x 46 m), the mixing height for the release is assumed to be 10 meters, and the average wind speed is assumed to 

be 2 m/s, the exposure time for any release would be less than 25 seconds. Using the dose 

coefficient for air submersion from current Federal guidance (EPA-402-R-93-081), the 

external dose from submersion in the atmospheric cloud would be more than three orders of 

magnitude lower than the dose from external exposure to 8

"Kr remaining in the waste pile. 

Therefore, external exposure to landfill workers to 8

"Kr released from the waste pile can be 

neglected. 

A.2.3.2.2 Inhalation Exposure to Individuals 

The annual EDE to an individual landfill worker from inhalation exposure is estimated using a 

model of the form given by Equation (2) in Appendix A.2.3.1.2. The mass of waste disposed 

annually in a landfill, Mw, and the exposure time, T, were the values given above for external

A.2-10
exposure. As described with Equation (2), the respirable fraction of airborne material, fa, was 

0.7; the breathing rate, Ua , was 1.2 m3

/h; and the inhalation dose coefficients for radionuclides, 

Dih, were obtained from current Federal guidance. The assumed atmospheric mass loading 

of waste, La, was 2x10 4 g/m3, which is a recommended value for the dust loading due to 

mechanical disturbance (Healy, 1980). 

A.2.3.2.3 Ingestion Exposure to Individuals 

The annual EDE to an individual landfill worker from ingestion exposure is estimated using a 

model of the form given by Equation (3) in Appendix A.2.3.1.3. The mass of waste disposed 

annually in a landfill, M,, and the exposure time, T, were the values given above for external 

and inhalation exposure. The ingestion dose coefficients for radionuclides, Ding, were obtained 

from current Federal guidance. The assumed ingestion rate of waste, Uig, was 0.06 g/h, 

which is a value appropriate for construction activities (EPA, OSWER Directive 9285.6-03). 

A.2.3.2.4 Collective Dose for Landfill Workers 

Based on discussions with a landfill operator (Phone call, J. Bailey, Browning-Ferris Industries, 

Knoxville, TN, July 1994), a total of five workers are assumed to be located at the open working 

face of a landfill during the working year of 1600 hours. Therefore, for any exposure pathway, 

the annual collective EDE to landfill workers is given by the annual individual dose multiplied by 

the factor 5 x 3,500 = 17,500, where 3,500 again is the assumed number of operating landfills 

(EPA-530-R-96-006). 

A.2.3.2.5 Results of Dose Calculations 

Tables A.2.3 and A.2.4 present the annual individual and collective EDEs to workers at 

municipal landfills from 1 year's disposals of a unit quantity of 1 YCi (37 kBq) of each 

radionuclide in all landfills estimated in this assessment (i.e., the DSRs in Equation (1)). In 

applying the DSRs to disposal of particular items containing radioactive material, the three 

corrections described in Appendix A.2.3.1.5 normally could be applied. The first is a reduction 

in all DSRs by a factor of 0.8 to take into account the fraction of all disposals sent to landfills 

when disposal by incineration also could occur. The second is a reduction in the DSRs for 

inhalation and ingestion that could be applied when the radioactive materials should be 

significantly less dispersible than loose materials in the waste. The correction factors for 

dispersibility discussed in Appendix A.2.3.1.5 for inhalation and ingestion exposure to waste 

collectors should be applicable to exposure to landfill workers. The third correction addresses 

the situation where there is limited distribution or use of the exempt material. As discussed in 

Section A.2.3.1.5, an adjustment should be applied to reflect localized use and disposal. If the 

assumed number of items to be disposed of annually is less than 3500, which is the assumed 

number of disposal sites, the DSRs should be increased by the ratio of 3500 divided by the 

number of items. 

A.2.3.3 Off-Site Members of the Public During Landfill Operations 

During operations at municipal landfills, off-site individuals and populations are assumed to be 

exposed to radionuclides released into the air and transported to off-site locations. The 

following four exposure pathways are assumed to occur: (1) inhalation of airborne 

radionuclides, (2) external exposure to airborne radionuclides, (3) external exposure to

A.2-11
radionuclides deposited on the ground surface, and (4) ingestion of food products contaminated 

by deposition onto the ground surface. 

For radionuclides in particulate form, the fraction of the material disposed released into the air 

during landfill operations is assumed to be 1x10-3. This release fraction is obtained from the 

following factors. First, the emission rate of particulate material during dumping and grading 

operations is assumed to be 4.3 kg/h (EPA-450/1-89-003). Second, this emission rate is 

reduced by a factor of 3 to account for the normal wetting of waste at a landfill to control 

airborne dust levels. Third, the emissions are assumed to occur for 2000 h/yr of operations. 

Finally, as described in Appendix A.2.3.2.1, the mass of waste disposed in a landfill annually is 

2.5x10 9 g. 

For 3H, which would not be in particulate form but is assumed to be present in soil water, the 

amount of disposed material that becomes airborne per year is estimated by multiplying the 

amount of 3

H per unit volume of waste after disposal by an assumed evapotranspiration rate of 

0.4 m/yr (Computer Codes, Yu et al., 1993) and by the assumed 2100 m2 area of the working 

face of the landfill from which evapotranspiration occurs. As discussed in Appendix A.2.3.2.1, 

half of the amount of 85

Kr disposed per year is assumed to become airborne during landfill 

operations. 

For the assumed releases of radionuclides into the air described above, annual individual and 

collective doses to nearby residents were calculated using CAP-88 (Computer Codes, Beres, 

1990). Assumptions used in the calculations are described as follows: 

The releases occur at ground level and uniformly over an assumed area of the working 

face at a landfill of 2100 M2 . The meteorological data (i.e., the distribution of wind speed 

and direction, annual rainfall, and average temperature) used in the calculations were 

for a site in Oak Ridge, TN. 

The rural agricultural data set contained in CAP-88 (Computer Codes, Beres, 1990) 

was used to evaluate the dose from ingestion of contaminated food products. For the 

vegetable, milk, and beef pathways, this data set specifies (1) the fraction of the 

ingested foodstuff produced at home, produced in the remainder of the assessment 

area, or imported, and (2) the density of beef and milk cattle and the fraction of the land 

surface cultivated for vegetable crops. 

In calculating collective dose, a population of 500,000 was assumed to be uniformly 

distributed within a distance of 80 km of each landfill. The assumed population was 

based on the average population density in the United States (Bureau of Census, 1990), 

and is intended to represent the variety of rural and semi-urban locations of landfills. 

Table A.2.5 presents the annual individual and collective EDEs to off-site residents due to 

airborne releases during operations at municipal landfills from 1 year's disposals of a unit 

quantity of 1 uCi (37 kBq) of each radionuclide in all landfills estimated in this assessment (i.e., 

the DSRs in Equation (1)). In applying the DSRs to disposal of particular items containing 

radioactive material, the three corrections described in Appendix A.2.3.1.5 normally could be 

applied. The first is a reduction in the DSRs by a factor of 0.8 to take into account the fraction 

of all disposals sent to landfills when disposal by incineration also could occur. The second is a 

reduction in the DSRs that could be applied when the radioactive materials should be

A.2-12
significantly less dispersible than loose materials in the waste. The latter reduction would be 

applied to all exposure pathways, because all of the pathways result from airborne releases. 

The correction factors for dispersibility discussed in Appendix A.2.3.1.5 for inhalation and 

ingestion exposure to waste collectors should be applicable to exposure to off-site residents. 

The third correction addresses the situation in which there is limited distribution or use of the 

exempt material. As discussed in Section A.2.3.1.5, an adjustment should be applied to reflect 

localized use and disposal. If the assumed number of items to be disposed of annually is less 

than 3500, which is the assumed number of disposal sites, the DSRs should be increased by 

the ratio of 3500 divided by the number of items. 

A.2.3.4 Off-Site Members of the Public Following Landfill Closure 

Following closure of a landfill, off-site individuals and populations are assumed to be exposed to 

radionuclides released into groundwater and transported to a nearby municipal well. Ingestion 

of drinking water obtained from the well is the only exposure pathway considered. 

For this scenario, the following distinction is made in estimating individual and collective doses. 

The estimated individual dose is the highest dose that would occur in any year from all 

disposals over the operating lifetime of a landfill, and the intent is to estimate the individual dose 

from actual disposal practices at a landfill. However, the estimated collective dose is the dose 

that would occur in an exposed population, over an assumed period of time after disposal, from 

only 1 year's disposals. The intent is to estimate the collective dose from the disposal of 

1 year's distribution of exempted items, based on an assumption that the number of items 

disposed per year would equal the annual distribution, for comparison with the annual collective 

dose during distribution and transport and routine use. The assessments of individual and 

collective dose for this scenario are described in the following paragraphs. 

A.2.3.4.1 Individual Dose From Well Water Use 

The annual EDE to an off-site individual who ingests water obtained from a well located near 

the landfill is estimated using a simple model for release of radionuclides into groundwater and 

transport to the well. Conceptually, a first-order leaching model is used to estimate the annual 

release of radionuclides into groundwater. The resulting concentration of radionuclides in 

groundwater is estimated by diluting the annual release in an assumed annual volume of 

groundwater flowing underneath the landfill, and the radionuclides are assumed to be 

transported in groundwater to the well without dispersion or further dilution. Therefore, the 

concentrations of radionuclides at the well are reduced relative to the concentrations in 

groundwater beneath the landfill only by radioactive decay during the travel time from the 

landfill to the well. 

Based on the simple conceptual model for release from a landfill and transport to a well 

described above, the annual EDE to an individual, H,, from consumption of radionuclide i in 

drinking water can be written as 

Hj (rem/yr) = (A T-, i x A L, Adq , x Uw x D ing, i x exp(-A R, i ti) , (5)

A.2-13
where A T, i = total activity of radionuclide i in landfill at time leaching begins (j-Ci), AL J = leaching constant from landfill into groundwater for radionuclide i(1/yr), 

qW= annual dilution volume of water beneath landfill (m3

/yr), U,, = ingestion rate of water from well by individual (m3

/yr), Ding, i = ingestion dose coefficient for radionuclide i (rem//gCi), 

A R'i = decay constant for radionuclide i (1/yr), and 

tj = travel time in groundwater from landfill to well for radionuclide i (yr). 

Leaching of radionuclides into groundwater is assumed to begin when disposal operations at 

the landfill cease, which maximizes the activity of radionuclides available for leaching. Then, if A i is the annual disposal of a unit activity of 1 pCi (37 kBq) of radionuclide i in all landfills, the total activity of radionuclide i in a single landfill at the time leaching begins is given by 

A T, = (1/NL) X (A /A R, ) [1 - exp(-A R, TL)], (6) 

where the factor 1/NL again represents the assumption described in Appendix A.2.3.1.1 that the annual disposals of radionuclides are distributed equally among all operating landfills, and the 

rest of this equation gives the activity at the end of the operating lifetime of the landfill, TL, taking into account the annual disposals, A, , and radioactive decay during the operating period. By combining Equations (5) and (6), the annual individual EDE from an annual disposal of the unit activity of radionuclide i is given by 

H,/A, (rem/llCi) = (1/NJ x (1/A R, ,) [1 - exp(-A R, ,TJ] x (A L, A/q ) X U (7) 

x Ding, ix exp(-A R, td). 

The leaching constant, A L, i, and the travel time in groundwater from the landfill to the well, tj, for radionuclide i in Equation (7) are described in the following paragraphs. 

The leaching constant for radionuclide 4 AL , i, is based on the model of Baes and Sharp (1983) for a saturated medium, corrected for leaching in an unsaturated medium (NUREG/CR-4370; Computer Codes, Rogers and Hung, 1987). The leaching model for a saturated medium has previously been used, for example, in a generic assessment for releases from contaminated 

soil (NUREG/CR-5512). The leaching constant is written as 

A L,, (Iyr) = (I/ed.R ,) x (I/Kj), (8) 

where I = infiltration rate of water through landfill (m/yr), 

e = volumetric water content of material in landfill (dimensionless), 

d,= thickness of waste in landfill (m), 

R; = retardation factor for transport of radionuclide i in water, and 

KS = saturated hydraulic conductivity of material in landfill (m/yr).

A.2-14
The factor I/edRi is the leaching constant for a saturated medium (Baes and Sharp, 1983), and the factor I/K,, called the contact time fraction (NUREG/CR-4370), is the correction for 

leaching in an unsaturated medium. The contact time fraction takes into account that leaching of radionuclides in an unsaturated medium occurs only during the fraction of the time that water 

is infiltrating through the medium. The retardation factor, R,, is given by (Baes and Sharp, 

1983) 

Ri = 1 + pKd.e, (9) 

where the parameter 8 is listed with Equation (8) above and 

p = bulk density of material in landfill (g/cm3), and 

Kd, i = solid/solution distribution coefficient for radionuclide i (mL/g). 

The travel time in groundwater from the landfill to the well, t,, for radionuclide i is given 

by 

tj = RiX/V,,, (10) 

where R; is the retardation factor for transport of radionuclide i in water in Equation (9) and 

X = distance of groundwater flow from edge of landfill to well (m), and 

V,, = groundwater velocity (m/yr). 

The model for estimating the annual individual dose from an annual disposal of a unit activity of 

a radionuclide in all landfills is given by Equations (7) to (10). The assumed values of 

parameters in Equation (7) that are not contained in other equations are described as follows: 

0 The operating lifetime of a landfill, TL, was 30 years, based on data for operating and 

closed facilities (EPA/530-SW88-034). 

* The annual dilution volume of water beneath the landfill, q,, was 7x10 4 m3

/yr, obtained 

as described below. 

0 The ingestion rate of water from the well by an individual, U,, was 0.73 m3/yr, based on 

a consumption rate of drinking water of 2 L/day (EPA/600/P-95/OO2Fa). 

* The ingestion dose coefficient, D g, for each radionuclide was obtained from current 

Federal guidance (EPA-520/1-88-020). 

The dilution volume for radionuclides leached from the landfill into groundwater was assumed to 

be the annual precipitation multiplied by the area of the landfill (EPA RAE-9232/1-2). As 

described later in this section, the annual infiltration of water through the landfill was assumed 

to be one-half of the annual precipitation. Therefore, the assumed dilution volume of water is 

twice the volume of water infiltrating through the landfill. This assumption is intended to be

A.2-15
representative of sites where only local recharge of groundwater occurs and the distance from 

the landfill to the well is comparable to or less than the dimension of the landfill parallel to the 

groundwater flow path. The assumed dilution volume would be quite conservative at sites 

where the amount of groundwater flowing beneath the landfill is much greater than the local 

recharge. In this assessment, the assumed precipitation was 0.9 m/yr, which is representative 

of sites with relatively abundant rainfall, and the assumed area of the landfill is 7.6x10' in'2, based on a typical size of planned facilities (EPA/530-SW88-034). The product of the annual 

precipitation and the area of the landfill gives the assumed dilution volume. 

The assumed values of the parameters in Equations (8) to (10) are described as follows: 

The infiltration rate of water through the landfill, /, was 0.45 m/yr, based on an 

assumption that infiltration is one-half of total precipitation (EPA RAE-9232/1-2), which 

is appropriate for sites with relatively abundant rainfall, and an assumed annual 

precipitation of 0.9 m/yr. 

* The volumetric water content of material in the landfill, e, was 0.3, which is a 

representative upper bound for different types of soil (Baes and Sharp, 1983). 

0 The thickness of waste in the landfill, d,,, was 10 meters, based on the assumed area 

of the landfill given above and typical waste volumes for a landfill 

(EPA/530-SW88-034). 

0 The bulk density, p, of material in the landfill was 1.4 g/cm 3 (Baes and Sharp, 1983). 

* The solid and solution distribution coefficient, Kd, for each radionuclide was the value 

adopted in NUREG/CR-5512, and is given in Table A.2.6. 

0 The contact time fraction, I/Ks, for waste in the unsaturated zone was 5xl 0-3, as 

described below. 

0 The distance of groundwater flow from the edge of the landfill to the well, X, was 

100 meters, based on an assumption that the well is located at the boundary of the 

buffer zone. 

a The groundwater velocity, V,, was 10 m/yr, which is representative of sites with 

relatively fast groundwater flow (NUREG/CR-4370). 

The assumed contact time fraction for waste in the unsaturated zone was based on values 

developed by the NRC for reference sites in different regions of the United States 

(NUREG/CR-4370), and is intended to represent an average value at sites with relatively 

abundant rainfall. For an assumed annual infiltration of water of 0.45 m/yr, the assumed 

contact time fraction of 5x1 0-3 corresponds to a saturated hydraulic conductivity, K,, in the 

landfill of about 1xi 02 m/yr. This is toward the low end of representative values for different soil 

types (Computer Codes, Yu et al., 1993) and, thus, would result in somewhat conservative 

values of the contact time fraction at many sites. 

In estimating the annual individual dose using Equations (7) to (10), an additional constraint is 

applied in regard to the time period of concern. In particular, an individual dose is calculated for

A.2-16
a radionuclide only if the travel time from the landfill to the well obtained from Equation (10) is 

1000 years or less. The assumed time period for the calculations is based on the NRC's stated 

intention that its radiological criteria for decontamination and decommissioning of contaminated 

sites would be applied only for 1000 years (NRC, 62 FR 39058). Therefore, for the assumed 

distance from the landfill to the well of 100 meters and the assumed groundwater velocity of 

10 m/yr, an individual dose is calculated for a radionuclide only if the retardation factor is 100 or 

less (i.e., if the distribution coefficient in Table A.2.6 is about 21 mL/g or less). 

A.2.3.4.2 Collective Dose from Well Water Use 

As discussed at the beginning of Appendix A.2.3.4, the collective dose for releases from a 

landfill to groundwater is calculated for 1 year's disposal of a unit activity of each radionuclide, 

rather than the annual disposals over the operating lifetime of the landfill. In this assessment, 

the disposals are assumed to occur in the last year of operations, which maximizes the 

collective dose. Therefore, based on Equations (5) and (7) for the annual individual dose from 

disposals over the operating lifetime of landfills, the annual individual dose from 1 year's 

disposals used in the calculation of collective dose is given by 

H, (rem/yr) = (1C) x (1/NL) x (A L, Jq,) x Uw x Ding, ix exp(-ARjti). (11) 

The calculation of collective dose from use of well water near landfills is based on the annual 

individual dose in Equation (11) and the following assumptions. First, the population served by 

all municipal wells located near landfills is 700,000 (EPA/530-SW88-034); i.e., the average 

population at each of the 3,500 landfills is 200. Second, the collective dose is calculated for the 

time period between the arrival of the contaminant plume, as obtained from Equation (10), and 

1000 years, which is the time period of concern for the calculations discussed in the previous 

section. Thus, the collective dose over 1,000 years from 1 year's disposals of a unit quantity of 

a radionuclide in all landfills is obtained by integrating the collective dose for the first year of 

exposure, as obtained from the annual individual dose in Equation (11) and the assumed 

population of 700,000, from time t, in Equation (10) to 1,000 years, taking into account 

radioactive decay over that time. 

A.2.3.4.3 Results of Dose Calculations 

For releases to groundwater and transport to a nearby well, the individual and collective EDEs 

to off-site residents from disposals of a unit quantity of 1 MCi (37 kBq) of each radionuclide in all 

landfills estimated in this assessment (i.e., the DSRs in Equation (1)) are presented in 

Table A.2.7. As discussed previously, the individual doses represent the annual dose from 

disposal of the unit quantity of each radionuclide during each year over the assumed 30-year 

operating lifetime of landfills, but the collective doses represent the dose over 1000 years from 

disposal of the unit quantity of each radionuclide during the last year of operations only. 

In applying the DSRs to disposal of particular items containing radioactive material, the 

following points should be noted. First, as in the other exposure scenarios for disposal in 

landfills, all DSRs normally should be reduced by a factor of 0.8 to take into account the fraction 

of all disposals sent to landfills when disposal by incineration also could occur.

A.2-17
Second, the DSRs for releases to groundwater assume the radioactive materials are dispersed in the landfill in a form that would be readily accessible to infiltrating water, but this would not be 

the case for many items. Therefore, depending on the physical form of the particular item of 

concern, the assessor could reduce the DSRs whenever the radioactive materials should be 

significantly less accessible to infiltrating water than finely dispersed materials. Based on the 

accessibility index for disposal of low-level radioactive waste developed by the NRC 

(NUREG-0782), the following guidelines for reducing the DSRs for releases to groundwater are 

suggested: 

For materials in the form of small bulk solids that should be significantly less accessible 

to infiltrating water than loose materials in waste or very small items, the DSRs may be 

reduced by a factor of 10. 

For large solid items, such as solid metal forms, that should be accessible to infiltrating 

water only at the surface of the waste form and should have a low leachability, the DSRs 

may be reduced by a factor of 100. 

The reduction factor for leaching of relatively inaccessible materials in water is similar in 

concept to the reduction factor applied to releases of less dispersible materials during waste 

collection and landfill operations that is discussed in Appendix A.2.3.1.5. However, during 

waste collection and landfill operations, releases from some types of items may be assumed to 

be zero (0), but an assumption of zero (0) release would not be reasonable for disposal, because even large metal, glass, or ceramic waste forms would be subject to leaching at the 

surface. 

An additional correction is needed for addressing the situation in which there is limited 

distribution or use of the exempt material. As discussed in Section A.2.3.1.5, an adjustment should be applied to reflect localized use and disposal. If the assumed number of items to be 

disposed of annually is less than 3500, which is the assumed number of disposal sites, the 

DSRs should be increased by the ratio of 3500 divided by the number of items. 

A.2.3.5 Future On-Site Residents at Landfills 

In the course of developing a model for exposure to future on-site residents at municipal landfill 

sites for use in the generic disposal methodology in Appendix A.2, two issues required consideration. The first was the types of exposure scenarios involving access to municipal landfill sites by members of the public that would be reasonable to assume following closure of 

a site and its release for public use. The second issue was the particular exposure pathways 

that should be assumed for the chosen exposure scenario. 

Based on available information, it appeared that the most common uses of municipal landfill 

sites following closure and release to the public would be as golf courses, public parks, or other 

recreational areas, or perhaps as an industrial park. This is reasonable when one considers 

that municipal landfills now are constructed, operated, and closed under Subtitle D of the 

Resource Conservation and Recovery Act (RCRA) in much the same way as hazardous waste 

disposal facilities and it is considered desirable not to unduly disturb disposed waste after 

closure. None of the expected uses of closed landfill sites involve permanent occupancy by members of the public. However, construction of housing at the sites is a credible, albeit

A.2-18
somewhat unlikely, occurrence and permanent residence in housing should result in higher doses to members of the public than the other credible uses noted above. 

Since the assumption of permanent on-site residence in housing already is expected to be 

conservative compared with more likely exposure scenarios at landfill sites, it was decided that 

exposure pathways should be included in the scenario only if they would be reasonably likely to 

occur. Based on this consideration, it was decided not to include a food ingestion (vegetable) pathway in the dose assessment for future on-site residents, because it is obvious from 

observation that most home owners do not have a vegetable garden, especially a garden at the 

same location as their home. In contrast, external exposure and intakes by inhalation of 

suspended activity or ingestion of waste materials either are unavoidable or are reasonably likely occurrences at any site and for any living habits, and these scenarios were included in the 

dose assessment for future on-site residents. 

The decision not to include a vegetable pathway in the dose assessment for future on-site 

residents is in accordance with recommendations of the International Commission on 

Radiological Protection (ICRP) in regard to the definition of critical groups to be used in 

radiation protection, as described, for example, in ICRP Publication 26 (see references). The 

critical group essentially is the population group expected to receive the highest doses, but the 

ICRP intends that doses should be assessed for average exposures within the critical group, rather than the maximum possible exposure to any individual. Thus, in the case of interest 

here, the critical population group consists of individuals who are assumed to reside on a 

municipal landfill site, and average (i.e., expected) doses to these individuals should be 

calculated. Since exposure to individuals due to consumption of vegetables grown on a landfill 

site is not expected to be a normal occurrence, the ICRP does not intend that this pathway should be included in a dose assessment for this critical group. Based on these considerations, 

the following modeling assumptions were used for estimating doses for future on-site residents 

at landfills. 

At some time following closure of a municipal landfill, members of the public are assumed to 

establish permanent residency in a suburban housing development on the landfill site. The 

following three exposure pathways are assumed to occur: (1) external exposure to 

radionuclides in the landfill during indoor and outdoor residence on the site, (2) inhalation of 

radionuclides suspended from the landfill into the air during indoor and outdoor residence on 

the site, and (3) ingestion of radionuclides in the waste. The existence of these exposure pathways is based on an assumption that waste in the landfill is uncovered during excavation of 

the site and remains uncovered during site occupancy. 

For this exposure scenario, the same distinction is made in estimating individual and collective 

doses as is made in the scenario for releases to groundwater described at the beginning of 

Appendix A.2.3.4. That is, the estimated individual dose is the highest dose that would occur in 

any year, due to all disposals over the operating lifetime of a landfill. For any radionuclide, this 

dose would occur at the time residence on the landfill site first occurs. However, the estimated 

collective dose is the dose that would occur in an assumed population, over an assumed time 

period after disposal, due only to 1 year's disposals during the last year of landfill operations. The dose assessment for future on-site residents is described in the following paragraphs.

A.2-19
A.2.3.5.1 External Exposure to Individuals

The annual EDE to an individual on-site resident from external exposure is estimated by assuming that the source is an infinitely thick, uniformly contaminated volume of soil. Using 

Equation (4) in Appendix A.2.3.2.1 and taking into account the buildup and decay of 

radionuclides disposed over the operating lifetime of the landfill, TL, as in Equation (6), the 

annual individual dose, Hi, from an annual disposal of a unit activity, A,, of 1 /Ci (37 kBq) of 

each photon-emitting radionuclide i in all landfills is given by 

H,/A , (reMlnCi) = (1/NL) x (1/A R ,)d[1 - exp(-A R, TL)] x (1/M. TL) (12) 

x [(f , x fs) + four] x Oext, i x exp(-A R, i Tc), 

where the various parameters are defined with Equations (4) to (6), except the fraction of the 

year during which exposure occurs is separated into the fraction of the time indoors, f.,,, and 

outdoors, foWr, the shielding factor, fh, applies only during indoor residence, and the last term 

in this equation represents radioactive decay during the time, Tc, between closure of the 

facility and the establishment of permanent residency on the landfill site. 

The assumed number of operating landfills, NO the values of the mass of waste disposed in a 

landfill annually, M,, and the external dose coefficient, Dext, for each radionuclide are 

described with Equation (4), and the operating lifetime of the landfill, TL , again is assumed to 

be 30 years. The assumed values of the other parameters in Equation (12) are described as 

follows: 

The fraction of the year during which indoor exposure occurs, fi,, was 0.65 

(EPA/600/P-95/OO2Fa). 

The shielding factor during indoor residence, fSh, was 0.7 (NRC, Regulatory Guide. 

1.109). 

The fraction of the year during which outdoor exposure occurs, fot, was 0.05 

(EPA/600/P-95/002Fa). 

The time period between closure of the landfill and the establishment of permanent 

residency on the site, Tc, was 30 years. 

Given the assumptions about indoor and outdoor exposure times and the shielding factor during indoor residence, the contribution to external dose while outdoors is only about 10% and, thus, 

can be neglected. The assumed value of T. is based on the presumption that, given current 

requirements in 40 CFR 258 for post-closure activities at landfills under RCRA, the sites will not 

be released for unrestricted use by the public immediately upon closure. 

For 8

"Kr, the assessment of external dose also assumes that only half of the disposed activity remains in the waste following landfill operations (see Appendix A.2.3.2.1), and the dose 

obtained from Equation (12) is reduced by a factor of 2.

A.2-20
A.2.3.5.2 Inhalation Exposure to Individuals

The annual EDE to an individual on-site resident from inhalation exposure is estimated using a 

model of the form given by Equation (2) in Appendix A.2.3.1.2. Therefore, taking into account 

the buildup and decay of radionuclides disposed over the operating lifetime of the landfill, TL, as 

in Equation (6), and the time between landfill closure and the onset of permanent residency, Tc, 

as in Equation (12), the annual individual dose, H,, from an annual disposal of a unit activity, Ai 

of 1 gCi (37 kBq) of each radionuclide i in all landfills is given by 

H,/A, (reml/Ci) = (1/NL) x (1/AR, j) [1 - exp(-AR, iTL)]x (1/M, TL) X fa X Ua (13) 

x [ff x L., in + (fox, x x Djh, ix exp(-AR, TC), 

where the various parameters are defined with Equations (2), (4), (5), (6), and (12), except U, 

is the annual breathing rate and the atmospheric mass loading is separated into values indoors, 

La, ,and outdoors, Laout,which are applied to the corresponding indoor and outdoor exposure 

times. 

The assumed number of operating landfills, NL, and the values of the mass of waste disposed in a landfill annually, M,, the respirable fraction of airborne material, f., and the inhalation 

dose coefficient, Dinh, for each radionuclide are described with Equations (2) and (4). The 

operating lifetime of the landfill, TL, and the time delay before the onset of permanent residency, Tc, each are again assumed to be 30 years. The fraction of the year during which 

exposure occurs indoors, f;,, and outdoors, four, again are assumed to be 0.65 and 0.05, 

respectively. Assumed values of the other parameters in Equation (13) are described as 

follows: 

The annual breathing rate, Ua , was 8400 m3/yr, based on an assumed breathing rate 

for resting and light activity of 23 m3/day (EPA/600/P-95/002Fa). 

The atmospheric mass loading of waste outdoors, La, aou, was 4x10' g/m 3

, which is an 

average background dust loading (Anspaugh et al., 1975). 

The atmospheric mass loading of waste indoors, La, ,, was one-third of the value 

outdoors (EPA, RAE-9232/1-2). 

For thorium, the dose from inhalation exposure considers the contribution from 2°Rn during indoor residence. The dose from exposure to 2°Rn during indoor residence is obtained from a 

natural analog model (Cook and Hunt, 1994), which is based on the known average dose from 

indoor radon per unit concentration of the parent radionuclide in surface soil. For an indoor 

residence time of 0.5, the natural analog model gives an EDE from exposure to 22°Rn of 

2.7 sievert (Sv)/yr per GBq/m3 (1.Oxl 0-2 rem/yr per UCi/m 3) of 232Th in soil. Thus, for the indoor 

residence time of 0.65 assumed in this assessment, the EDE from inhalation of '°Rn is 

3.5 Sv/yr per GBq/m 3 (1.3x1 0-2 rem/yr per /,Ci/m3) of 232Th in soil. The dose from exposure to 

"•°Rn during outdoor residence is only a few percent of the dose during indoor residence (Cook 

and Hunt, 1994) and, thus, can be neglected.

A.2-21
For uranium, only the short-lived decay products that would be in activity equilibrium with the 

parent uranium isotopes at times shortly after chemical separation are considered in this 

assessment. Therefore, the dose from inhalation of 'Rn is not considered, because the 

parent radionuclide "

6 Ra builds up in the waste only at times long after chemical separation. 

A.2.3.5.3 Ingestion Exposure to Individuals 

The annual EDE to an individual on-site resident from ingestion exposure is estimated using a 

model of the form given by Equation (3) in Appendix A.2.3.1.3. Therefore, taking into account 

the buildup and decay of radionuclides disposed over the operating lifetime of the landfill, TL, 

as in Equation (6), and the time between landfill closure and the onset of permanent residency, 

Tc, as in Equation (12), the annual individual dose, Hi, from an annual disposal of a unit 

activity, A ;, of 1 ptCi (37 kBq) of each radionuclide i in all landfills is given by 

H1/A j (rem/fiCi) = (1/NL) x (/A R, ) [1 - exp(-A R, i TL) x (1/Mw TL) X Ung x T (14) 

x Dig, j x exp(-A R, i Tc), 

where the various parameters are defined with Equations (2) to (6) and (12). 

The assumed number of operating landfills, NL, and the values of the mass of waste disposed 

in a landfill annually, M,, and the ingestion dose coefficient, Ding, for each radionuclide are 

described with Equations (2) to (4). The operating lifetime of the landfill, TL , and the time delay 

before onset of permanent residence, Tc, each are again assumed to be 30 years. Assumed 

values of the other parameters in Equation (14) are described as follows: 

The ingestion rate of waste, Uing, was 4x10 3 g/h, which is a value appropriate for 

residential ingestion of soil and dust (EPA, OSWER Directive 9285.6-03). 

The exposure time was 440 h/yr, based on an assumption that ingestion exposure 

occurs mainly during outdoor residence on the site and that the fraction of the year 

during which exposure occurs outdoors is 0.05 (EPA/600/P-95/OO2Fa). 

A.2.3.5.4 Collective Dose for Future On-Site Residents 

As discussed at the beginning of Appendix A.2.3.5, the collective dose for future on-site 

residents at a landfill is calculated for 1 year's disposals of a unit activity of each radionuclide, 

rather than the annual disposals over the operating lifetime of the landfill. In this assessment, 

the disposals are assumed to occur in the last year of operations. Therefore, the annual 

individual dose from 1 year's disposals used in the calculation of collective dose is obtained 

from Equations (12) to (14) by omitting the term describing the buildup and decay of activity 

during the operating lifetime of the landfill, TL. This approach essentially distributes the year's 

disposals over the entire landfill. 

The calculation of collective dose for future on-site residents at landfills is based on the 

individual dose calculated as described above and the following assumptions. First, the number 

of residents at each landfill site, based on the average density of suburban populations in the

A.2-22
United States (Bureau of Census, 1990) and the size of a typical landfill (EPA/530-SW88-034), 

is 10 (i.e., the total number of residents at all 3,500 landfill sites is 35,000). Second, as 

described in Appendix A.2.3.4.2, the collective dose is calculated by integrating the individual 

dose over 1000 years, taking into account the exposed population and radioactive decay over 

that time. 

A.2.3.5.5 Results of Dose Calculations 

Tables A.2.8 and A.2.9 present the annual individual and collective EDEs to future on-site 

residents from disposals of a unit quantity of 1 u.Ci (37 kBq) of each radionuclide in all landfills 

estimated in this assessment (i.e., the DSRs in Equation (1)). As discussed previously, the 

individual doses in Table A.2.8 represent the annual dose from disposal of the unit quantity of 

each radionuclide during each year over the assumed 30-year operating lifetime of landfills, but 

the collective doses in Table A.2.9 represent the dose over 1000 years from disposal of the unit 

quantity of each radionuclide during the last year of operations only. 

In applying the DSRs to disposal of particular items containing radioactive material, the 

following points should be noted. First, as in the other exposure scenarios for disposal in 

landfills, all DSRs normally should be reduced by a factor of 0.8 to take into account the fraction 

of all disposals sent to landfills when disposal by incineration also could occur. 

Second, the DSRs for inhalation and ingestion assume that the radioactive materials are 

dispersed in the landfill in a form that would be readily suspended into the air or ingested, but 

this would not be the case for many items. Therefore, depending on the physical form of the 

particular item of concern, the assessor could reduce the DSRs for these exposure pathways 

whenever the radioactive materials should be significantly less dispersible than loose materials 

in the waste. Because exposures are assumed to occur well after landfill closure, the correction 

factors for leachability in water described in Appendix A.2.3.4.3 should be appropriate for 

inhalation and ingestion exposure to future on-site residents, rather than the correction factors 

for dispersibility during landfill operations described in Appendix A.2.3.1.5. 

An additional correction is needed for addressing the situation where there is limited distribution 

or use of the exempt material. As discussed in Section A.2.3.1.5, an adjustment should be 

applied to reflect localized use and disposal. If the assumed number of items to be disposed of 

annually is less than 3500, which is the assumed number of disposal sites, the DSRs should be 

increased by the ratio of 3500 divided by the number of items. 

A.2.4 Dose Assessment for Disposal in Incinerators 

This section presents the models and parameter values for estimating individual and collective 

doses from disposal of radioactive materials in municipal incinerators, and the results of the 

dose assessment in the form of doses per unit activity of radionuclides disposed in all 

incinerators (i.e., the DSRs in Equation (1)) are tabulated. The groups of individuals considered 

in the dose assessment and their associated exposure pathways are described in 

Appendix A.2.2.2.

A.2-23
A.2.4.1 Waste Collectors

The exposure scenario and exposure pathways for waste collectors at municipal incinerators 

would be the same as for waste collectors at landfills. Therefore, except for the assumption 

about the number of operating facilities, the models and parameter values for estimating 

individual and collective doses for waste collectors at incinerators would be the same as those 

presented in Appendix A.2.3.1. 

In this assessment, the number of operating incinerators is assumed to be 150 (Kiser, 1995). 

Therefore, since the dose to individual waste collectors from external, inhalation, and ingestion 

exposure per unit activity of radionuclides disposed in all incinerators is inversely proportional to 

the number of incinerators (see Appendix A.2.3.1), the individual dose would be a factor of 

3500/150 = 23 higher than the corresponding individual dose for waste collectors for the same 

unit activity disposed at all municipal landfills, where 3500 is the assumed number of operating 

landfills. However, the collective dose to all waste collectors at incinerators per unit activity disposed at all incinerators would be the same as the collective dose to all waste collectors at 

landfills for the same unit activity disposed at all landfills. 

Tables A.2.10 and A.2.1 1 present the annual individual and collective EDEs to waste collectors 

at municipal incinerators from 1 year's disposals of a unit quantity of 1 gCi (37 kBq) of each 

radionuclide in all incinerators estimated in this assessment (i.e., the DSRs in Equation (1)). In 

applying the DSRs to disposal of particular items containing radioactive material, the following 

points should be noted. 

First, as discussed in Appendix A.2.3.1.5, if a particular item could be sent to either landfills or 

incinerators for disposal, the DSRs should be reduced by the fraction of the total annual 

disposals assumed to be sent to incinerators. Recent data indicate that the amount of waste 

sent to incinerators is about one-fourth of the amount sent to landfills (EPA-530-R-94-042). 

Therefore, if the input to the dose assessment is an assumed total activity of a radionuclide 

disposed per year in all landfills and incinerators, the DSRs for all exposure pathways should be 

reduced by a factor of 0.2 to consider the fraction of the disposed activity sent to incinerators. 

Second, for some items, such as large bulk metal forms that are not normally used in consumer 

products, it may be reasonable to assume that none of the material would be sent to 

incinerators for disposal. In these cases, the exposure scenario could be assumed not to apply. 

Also, if the number of items to be incinerated annually is less than 150, which is the number of 

assumed incinerators, the DSRs should be increased by the ratio of 150 divided by the number 

of items. 

Third, the DSRs for inhalation and ingestion exposure could be reduced whenever the 

exempted items of concern should be significantly less dispersible than loose materials in the 

waste. Suitable correction factors for dispersibility are discussed in Appendix A.2.3.1.5. 

A final correction addresses the situation in which there is limited distribution or use of the 

exempt material. As discussed in Section A.2.3.1.5, an adjustment should be applied to reflect 

localized use and disposal. If the assumed number of items to be disposed of annually is less 

than 3500, which is the assumed number of disposal sites, the DSRs should be increased by 

the ratio of 3500 divided by the number of items.

A.2-24
A.2.4.2 Workers at Incinerator

Workers at municipal incinerators are assumed to receive external, inhalation, and ingestion 

exposures while located at the edge of a partially enclosed waste pit where garbage trucks 

unload waste at the facility. The dose assessment for incinerator workers is described in the 

following paragraphs. 

A.2.4.2.1 External Exposure to Individuals 

The annual EDE to an individual incinerator worker from external exposure is estimated by assuming that the source is an infinitely thick, uniformly contaminated volume of soil-equivalent 

material and that the worker is standing at the edge of the source volume. The individual dose, 

Hi, from an annual disposal of a unit activity, A,, of 1 4Ci (37 kBq) of each photon-emitting 

radionuclide i in all incinerators is estimated using a model of the form given by Equation (4) in 

Appendix A.2.3.2.1, with the following changes in the model and parameter values: 

0 The factor 1/NL, where NL is the number of operating landfills, is replaced by the factor 

1/N,, where N, is the number of operating incinerators which, as noted in 

Appendix A.2.4.1, is assumed to be 150. 

0 The mass of waste disposed in an incinerator annually, M,, was 2.0x10'1 g, as 

described below. 

a The fraction of the year during which exposure occurs, fex, was 0.09, based on an 

assumption that a worker spends 1600 h/yr in the vicinity of the waste pit but only half of 

this time at the edge of the waste pile. 

0 The shielding factor, f h, is unity for a worker standing at the edge of the waste pile. 

0 For a worker standing at the edge of the waste pile, the external dose coefficient (De,,,) 

for each radionuclide is one-half of the value for a source of infinite lateral extent. 

The mass of waste disposed in an incinerator was assumed to be the average value for all 

incinerators. It was calculated from the reported amount of municipal waste combusted in 1993 

of about 3.OxlO'0 kg (EPA-530-R-94-042) and the assumed number of incinerators. 

As described in Appendix A.2.3.2.1, external exposure to incinerator workers to 85Kr is 

estimated by assuming that half of the activity would be released into the air during unloading of 

waste, due to breakage of half of the containers for this radionuclide. Therefore, the external 

dose from 85

Kr retained in the waste would be half of the value calculated as described above. 

A.2.4.2.2 Inhalation Exposure to Individuals 

The annual EDE to an individual incinerator worker from inhalation exposure is estimated using 

a model of the form given by Equation (2) in Appendix A.2.3.1.2. The number of operating 

incinerators, N,, and the mass of waste disposed in an incinerator annually, M", were the 

values given above for external exposure. The respirable fraction of airborne material, fa, again was 0.7, and the assumed values for the other parameters that differ from the values 

given with Equation (2) are described as follows:

A.2-25
The atmospheric mass loading of waste, La , was 4x10-4 g/m 3 , based on measurements 

at an operating facility (Hahn et al., 1989). 

* The exposure time for the worker, T, was 1600 h/yr. 

A.2.4.2.3 Ingestion Exposure to Individuals 

The annual EDE to an individual incinerator worker from ingestion exposure is estimated using 

a model of the form given by Equation (3) in Appendix A.2.3.1.3. The number of operating incinerators, N,, the mass of waste disposed in an incinerator annually, M", and the exposure 

time for the worker, T, were the values given above for external and inhalation exposure. The 

assumed ingestion rate of waste, Uig, was 6x10-3 g/h, which is a value appropriate for 

commercial or industrial activities (EPA, OSWER Directive 9285.6-03). 

A.2.4.2.4 Collective Dose for Incinerator Workers 

In this assessment, a total of two workers are assumed to be located near the waste pit at each 

incinerator during the working year. Therefore, for any exposure pathway, the annual collective 

EDE to incinerator workers is given by the annual individual dose multiplied by the factor 

2 x 150 = 300, where 150 again is the assumed number of operating incinerators. 

A.2.4.2.5 Results of Dose Calculations 

Tables A.2.12 and A.2.13 present the annual individual and collective EDEs to workers at 

municipal incinerators from 1 year's disposals of a unit quantity of 1 UCi (37 kBq) of each 

radionuclide in all incinerators estimated in this assessment (i.e., the DSRs in Equation (1)). In 

applying the DSRs to disposal of particular items containing radioactive material, the 

corrections described in Appendix A.2.4.1 normally could be applied. The first is a reduction in 

all DSRs by a factor of 0.2 to take into account the fraction of all disposals sent to incinerators 

when disposal in landfills also could occur. Alternatively, when appropriate, it could be 

assumed that none of the items would be sent to an incinerator for disposal, and the exposure 

scenario could be assumed not to apply. The second correction is a reduction in the DSRs for 

inhalation and ingestion that could be applied when the radioactive materials should be 

significantly less dispersible than loose materials in the waste. The same correction factors for 

dispersibility discussed in Appendix A.2.3.1.5 should be applicable to inhalation and ingestion 

exposure to incinerator workers. Additionally, where there is limited distribution or use of the 

exempt material as discussed in Section A.2.3.1.5, an adjustment should be applied to reflect 

localized use and disposal. Also, if the number of items to be incinerated annually is less than 

150, which is the number of assumed incinerators, the DSRs should be increased by the ratio 

of 150 divided by the number of items. 

A.2.4.3 Off-Site Members of the Public 

During operations at incinerators, off-site individuals and populations are assumed to be 

exposed to radionuclides released into the air and transported to off-site locations. The 

following four exposure pathways are assumed to occur: (1) inhalation of airborne 

radionuclides, (2) external exposure to airborne radionuclides, (3) external exposure to 

radionuclides deposited on the ground surface, and (4) ingestion of food products contaminated 

by deposition onto the ground surface.

A.2-26
Airborne releases at incinerators are assumed to result entirely from waste incineration (see 

Appendix A.2.2.2.3). For radionuclides in particulate form, the fraction of material incinerated 

that is released into the air from the incinerator stack is assumed to be 1.9x1 0-4 

(EPA-450-AP-42-5ED). For 3H, which would not be in particulate form, all of the amount of 

material disposed is assumed to be released into the air during incineration. Finally, half of the 

total amount of 8

'Kr sent to incinerators is assumed to be released during operations at the 

waste pits (see Appendix A.2.4.2.1), and the other half is assumed to be released during incineration. However, for ease of calculation, doses are calculated by assuming that all of the 

85Kr sent to incinerators is released from the stack during incineration. 

For the assumed releases of radionuclides into the air described above, annual individual and 

collective doses to nearby residents were calculated using CAP-88 (Computer Codes, Beres, 

1990). Assumptions used in the calculations are described as follows: 

0 Releases occur at a height of 43 meters through a stack of diameter 1.5 meters. The 

source temperature was 1270C and the exit velocity was 7.3 m/s. The meteorological 

data used in the calculations were for a site in Islip, NY, which is in an area with a 

substantial number of operating incinerators. 

0 The urban agricultural data set in the CAP-88 computer code (see Appendix A.2.3.3) 

was used to evaluate dose from ingestion of contaminated food products. The choice of 

this data set reflects the location of many incinerators in urban or suburban areas. 

0 In calculating collective dose, a population of 2 million was assumed to be uniformly 

distributed within a distance of 80 km of each incinerator. The assumed population was 

based on the average population density in suburban areas in the United States (Bureau 

of Census, 1990), and is intended to represent the urban or suburban locations of many 

incinerators. 

Table A.2.14 presents the annual individual and collective EDEs to off-site residents due to 

airborne releases during waste incineration from 1 year's disposals of a unit quantity of 1 PCi 

(37 kBq) of each radionuclide in all incinerators estimated in this assessment (i.e., the DSRs in 

Equation (1)). In applying the DSRs to disposal of particular items containing radioactive 

material, the values normally should be reduced by a factor of 0.2 to take into account the 

fraction of all disposals sent to incinerators when disposal in landfills also could occur. 

Alternatively, when appropriate, it could be assumed that none of the items would be sent to an 

incinerator for disposal, and the exposure scenario could be assumed not to apply. 

Additionally, where there is limited distribution or use of the exempt material as discussed in 

Section A.2.3.1.5, an adjustment should be applied to reflect localized use and disposal. Also, 

if the number of items to be incinerated annually is less than 150, which is the number of 

assumed incinerators, the DSRs should be increased by the ratio of 150 divided by the number 

of items. 

A.2.5 Dose Assessment for Metal Recycling 

This assessment also considers doses to individual members of the public resulting from 

recovery of items containing exempted amounts of radioactive materials for use in recycled 

ferrous metals (e.g., steel). The groups of individuals considered in the dose assessment and

A.2-27
their associated exposure pathways are described in Appendix A.2.2.3. Only individual doses 

are estimated for this disposal option, because recycling in metals is expected to be an unusual 

occurrence for most exempted materials. 

The estimates of individual dose from recycling of radionuclides in metals obtained in this 

assessment are based directly on results of a previous study by Hill et al. (1995). The 

assessment methodology used by Hill et al. resembles the methodology developed previously 

by the International Atomic Energy Agency (IAEA) to derive exemption levels for recycle and 

reuse of materials containing trivial quantities of radionuclides (IAEA SS No. 111-P-1.1). 

Estimates of individual dose were obtained by Hill et al. (1995) for slag workers at metal 

smelters, users of recycled products, and members of the public residing near smelters. In this 

assessment, the previous results, which are in the form of annual EDEs per unit activity 

concentration of radionuclides in the feed material sent to a smelter, are converted to annual 

EDEs per unit activity of radionuclides sent to all smelters using the following assumptions. 

First, the number of facilities at which most smelting occurs is 100 (Phone call, C. Bechak, 

Steel Manufacturing Association, Washington, DC, June 1996), and the unit activity of 

radionuclides disposed per year is distributed equally among all smelters. Second, in 

estimating doses to slag workers and users of recycled products, about 100 Mg/yr of metal are 

sent to each smelter and are incorporated in the finished product, and about 10 Mg/yr of the 

feed material are incorporated in the slag (Hill et al., 1995). Third, in estimating doses to 

off-site members of the public, the fraction of radioactive material released to the air during 

smelting is assumed to be 6.5x10- 3 (EPA-450-AP-42-5ED), except all 3H is assumed to be 

released. 

In this assessment, the recycled material is assumed to be steel used in automobiles (Hill et al., 

1995). Tables A.2.15 and A.2.16 present the annual individual EDEs from 1 year's disposals of 

a unit quantity of 1 uCi (37 kBq) of each radionuclide in all smelters estimated in this 

assessment (i.e., the DSRs in Equation (1)). The results in Table A.2.15 give the estimated 

dose to an individual slag worker or user of an automobile, whichever is greater, and an 

identification of the critical exposure pathway. Table A.2.16 gives the estimated dose to an off

site member of the public from airborne emissions. The DSRs in these tables can be applied to 

assumed activities of radionuclides sent to all smelters annually to obtain estimates of annual 

EDEs to exposed individuals. 

A.2.6 Application of Results 

This section describes how the results given in this appendix are used to obtain estimates of 

individual and collective doses from disposal of exempt items containing radioactive material. 

The generic disposal methodology has been developed with the intention of minimizing the 

number of inputs and assumptions that the user must provide in applying the results. 

In many of the tables of results, separate DSRs are given for external, inhalation, and ingestion 

exposure. In general, the DSR for all exposure pathways is the sum of the DSRs for each 

pathway. 

Three types of inputs and assumptions must be provided by the user in applying the results in 

the tables. The first input, which is always required, is the assumed total quantity of a

A.2-28
radionuclide disposed annually in all disposal facilities. However, in estimating doses from 

disposal in landfills, incinerators, or metal smelters, no further assumptions are required about 

the total number of facilities receiving the assumed annual disposals, because such 

assumptions are incorporated in the model equations for calculating the DSRs given in the 

tables. 

The second input, which also is generally required, is an assumption about whether disposal in 

landfills and incinerators would occur, or whether disposal would occur only in landfills. In the 

latter case, the assumed annual disposals of a radionuclide would be applied directly to the 

DSRs in the tables for disposal in landfills to obtain estimates of individual and collective doses, and the DSRs in the tables for disposal in incinerators would be ignored. However, if disposal in both landfills and incinerators is assumed to occur, the DSRs in the tables for landfills should 

be reduced by a factor of 0.8, and the DSRs in the tables for incinerators should be reduced by 

a factor of 0.2. The assumed total annual disposals of a radionuclide in all facilities then would 

be applied to these results. This type of adjustment is not used in applying the DSRs in the 

tables for recycling. 

The third type of input consists of assumptions about the extent to which radionuclides could be 

dispersed during waste collection and waste operations at landfills or incinerators or released 

following disposal in landfills, compared with loose materials in the waste. Specifically, the 

DSRs for inhalation and ingestion exposure to waste collectors, workers at landfills or 

incinerators, and future on-site residents at landfills, and the DSRs for exposure to off-site 

residents at landfills due to airborne releases during landfill operations, can be reduced if the radioactive materials are considered to be significantly less dispersible than loose materials in 

trash. The DSRs for exposure to off-site residents at landfills due to releases to groundwater 

can be reduced if the radioactive materials are considered to be significantly less accessible to infiltrating water than loose materials in the waste. Suggested dose reduction factors for these 

exposure scenarios are discussed in Appendixes A.2.3.1.5 and A.2.3.4.3. If no assumptions about dose reduction factors for dispersibility or accessibility to water are made by the user, then the default dose estimates would be based only on the first two inputs described above. 

A specific example is considered: disposal of 1 million items, each containing 1 g of thorium. 

Disposal in both landfills and incinerators is assumed to occur, and the radioactive material is assumed to be in a physical form that is 10 times less dispersible and 10 times less accessible 

to water than loose materials in the waste. Based on these assumptions, the following estimates of dose are obtained: 

Waste collectors at landfills (Tables A.2.1 and A.2.2) 

Annual individual dose = (106 g) [(3.1x10'-' rem/g) + (0.1)(5.8x10'- 1 rem/g) 

+ (0.1)(2.8x10-11 rem/g)] (0.8) = 2.6x10-3 mSv (0.26 mrem) 

Annual collective dose = (106 g) [(1. lxl 0-6 p-rem/g) + (0.1)(2.Oxl 0- 7 p-rem/g) 

+ (0.1)(9.9x10- 8 p-rem/g)] (0.8) = 8.8xl 0-3 person-Sv (0.88 person-rem)

A.2-29
Workers at landfills (Tables A.2.3 and A.2.4) 

Annual individual dose = (106 g) [(2.8x10`1 remi/g) + (0.1)(6.3x10- 12 rem/g) 

+ (0.1)(6.0x10- 12 rem/g)] (0.8) = 2.3x10 4 mSv (0.023 mrem) 

Annual collective dose = (106 g) [(4.8x10- 7 p-rem/g) + (0.1)(1.1x10-7 p-rem/g) 

+ (0.1)(1.Ox10-7 p-rem/g)] (0.8) = 0.004 person-Sv (0.40 person-rem) 

Off-site residents at landfills due to airborne releases (Table A.2.5) 

Annual individual dose = (106 g) [(0.1)(3.3x10- 3 rem/g)] (0.8) = 2.6x10-7 mSv 

(2.6x10-5 mrem) 

Annual collective dose = (106 g) [(0.1)(2.0x10- 7 p-rem/g)] (0.8) = 

1.6xl 0-4 person-Sv (0.016 person-rem) 

Off-site residents at landfills due to releases to groundwater (Table A.2.7) 

Annual individual dose = (106 g) [(0.1)(2.4x10-15 rem/g)] (0.8) = 1.9x10-9 mSv 

(1.9x10- 7 mrem) Collective dose not calculated 

(exposures occur beyond 1000 years) 

Future on-site residents at landfills (Tables A.2.8 and A.2.9) 

Annual individual dose = (106 g) [(9.2x10' 1

1 rem/g) + (0.1)(2.3x10-1° rem/g) 

+ (0.1)(1.1x10-' 3 rem/g)] (0.8) = 9.6x10-4 mSv (0.096 mrem) 

Collective dose over 1000 years = (106 g) [(1 .1x10 4 p-rem/g) 

+ (0.1)(2.6xl 0-4 p-rem/g) + (0.1)(1.3x10-7 p-rem/g)] (0.8) 

= 1.1 person-Sv (110 person-rem) 

Waste collectors at incinerators (Tables A.2.10 and A.2.1 1) 

Annual individual dose = (106 g) [(7.2x10- 9 rem/g) + (0.1)(1.4x10-9 rem/g) 

+ (0.1)(6.6x10-1" rem/g)] (0.2) = 0.015 mSv (1.5 mrem) 

Annual collective dose = (106 g) [(1 .lx0-6 p-rem/g) + (0.1)(2.0x10-7 p-rem/g) 

+ (0.1)(9.9x10-8 p-rem/g)] (0.2) = 2.2x10-3 person-Sv (0.22 person-rem) 

Workers at incinerators (Tables A.2.12 and A.2.13) 

Annual individual dose = (106 g) [(5.4x10-' 2 rem/g) + (0.1)(3.7x10-12 rem/g) 

+ (0.1)(1.8x10- 13 rem/g)] (0.2) = 1.2x10-5 mSv (1.2x10-3 mrem) 

Annual collective dose = (106 g) [(1.6x10- 9 p-rem/g) + (0.1)(1.1x10-9 p-rem/g) 

+ (0.1)(5.4x10-1 p-rem/g)] (0.2) = (3.4x10-6 person-Sv) 

(3.4xl 0-4 person-rem)

A.2-30
Off-site residents at incinerators due to airborne releases (Table A.2.14)

Annual individual dose = (106 g)(3.8x1 0-15 rem/g)(0.2) = 7.6x10 9 mSv 

(7.6x10-7 mrem) 

Annual collective dose = (106 g)(2.4x10-8 p-rem/g)(0.2) = (4.8x10-5 person-Sv) 

(4.8xl 0-3 person-rem) 

Thus, based on the generic disposal methodology, the highest annual individual EDEs would be 

(2.6 ,uSv (0.26 mrem)) to waste collectors at landfills or 0.015 mSv (1.5 mrem) to waste 

collectors at incinerators. The collective EDE from 1 year's distribution of the exempted items 

would be 1.1 person-Sv (110 person-rem), due almost entirely to exposure to future on-site 

residents at landfills for 1000 years after loss of institutional controls over the sites. If exposure 

to future on-site residents were not considered, the collective EDE would be 0.015 person-Sv 

(1.5 person-rem). 

If the same quantity of thorium were sent to smelters for metal recycling, the estimated annual 

individual EDEs, as obtained from Table A.2.15 and A.2.16, would be 0.48 mSv (48 mrem) to a 

slag worker and 6.3x10-6 mSv (6.3x10-4 mrem) to an off-site member of the public. However, 

since these dose values are less than lx1 0` mSv (<0.001 mrem) they would have been 

reported as "less than values" (i.e., less than 1x10-5 mSv (<0.001 mrem)) if they had been 

included in a specific assessment presented in this report. This is consistent with the notation 

that was used for this report.

A.2-31
Table A.2.1 Individual Dose-to-Source Ratios for Exposure to Waste 

Collectors at Municipal LandfillIS a

Radionuclide

14C 

36CI 

46

SC 

55 Fe 

6000 

63

Ni 

85Kr 

'Sr +'O 

99,TC 

1

106Ru + 106 Rh 

109

Cd + lo9

mAg 

1291 

1

33 Ba 

,, + 1

3 7m Ba 

147 Pm 

152 Eu 

204 T1 

2 10m Bi + 206

TI 

Th (natural)d 

U (natural)' 

U (depleted)' 

241

Am

External ExD~o

1.5x109

' 

1 .8X10-,

Individual Effective Dose Equivalent 

(rem/lgCi)b 

sure Inhalation Exposure' Ingest 

2.6x10'-7 

5.9x 0-'6 

6.2x 0-15 1 

8.3A10-15 

7.5x100-16 

6.1 X10-14 

8.7xl 0- 16

ion Exposure' 

3.3x1 0-15 

l.1X10-13 

I.6x100- 3 

3.3x1 Q-13 

3.2x1 0-14 

I.x1 Q-12 

~.X10- Q14

1 .9X1 0-12

6.6x10'-6 

1 .7A10` 0 

7.9x1 Q12 

1.1Xl0-,11 

3.4x1 QlO 

4.6x 10` 0 

4.Oxl Q-1 

8.9x10-` 

1 .4x10'2 

2.5x1 Q0` 

6.7x10'5 

9.4x-10-1O 

(2.Ux10-10 rem/g)e 

1.5x10 ,1 

(9.8x 0-12 rem/g)g 

2.2x101 ' 

(8.Oxl Ql2 remlg)' 

2.7x 10-11

3.7x1 Q13 

2.8x1 Q-16 

1.3x 0-13 

3.2A1 -14 

4.9x 0-14 

2.2x1 Q15 

9.OXl -15 

1 .1 X10-14 

6.2x1 0-14 

6.7A1 -16 

2.3A10-13 

2.6A1 -

12 

2.7x10-10 

(5.8x10 11 remlg)e 

3.5x10'1 

(2.4x1 Q11 rem/g)g 

3.4x1 Q11 

1 .2xl1010

8.Oxl Q-12 

7.6x1 0-14 

1 .4x1 Q-12 

6.8x1 Q13 

1.4x10'1 

1.8x10'3 

2.6x1 Q-12 

5.5X10Q14 

3.4x1 0-13 

1.8X10-13 

5.Oxl Q-12 

9.9X 10 Q1 

1 .3x10-'0 

(2.8x10-11 reMlg)e 

1.4x10' 11 

(9.9x 10 "1 remlg)g 

1.4x1 0-1 

(5.2x10-12 remig)' 

1 .9x1 010

See following page for footnotes.

A.2-32

3

H
Footnotes to Table A.2.1

a Models and parameter values are presented in Appendix A.2.3.1. Except as noted for thorium 

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from the 

disposal of a unit activity of 1 tCi (37 kBq) of radionuclides per year in all landfills. DSRs for all 

exposure pathways should be reduced by a factor of 0.8 if a unit activity of 1 uCi (37 kBq) of 

radionuclides per year is assumed to be disposed in landfills plus incinerators (see 

Appendix A.2.3.1.5). b 1 rem/izCi = 0.27 Sv/Bq. 

' Depending on the physical form of the particular items of concern, DSR ratios for the 

exposure pathway could be reduced whenever radioactive materials should be significantly less 

dispersible than loose materials in trash (see Appendix A.2.3.1.5). 

I Calculated DSRs take into account the contributions from the decay products of 232Th, which 

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and 

activity abundances of 232Th and 228

Th in natural thorium are given in Table 3.1.1. e Value in units of dose per unit mass is based on assumed activity abundances of 232

Th and 

22Th and the specific activity of 23"Th given in Table 3.1.3 (1 yCi = 4.59 g). 

'Calculated DSRs take into account the contributions from the short-lived decay products of 2

3U and 235

U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); 

assumed mass and activity abundances of 2•U, 235

U, and 2

3U in natural uranium are given in 

Table 3.1.1. 

9 Value in units of dose per unit mass is based on the assumed activity abundances of 238

U, 23

1U, and 2

'U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 jCi = 1.46 g). 

" Calculated DSRs take into account the contributions from the short-lived decay products of 

238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); 

assumed mass and activity abundances of 23U, 2

3

5U, and 23U in depleted uranium are given in 

Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 238U, 21U, and 21U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 yCi = 2.7 g).

A.2-33
apionuopaH

. sillipue-1 ledmunW le sioloalloo 

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odx3 juuj9jx3

'Seloulool jol 96ed buimoliol 99S

L-OLX9.9 

,(l5/w;aj-dg-OLXq*0 

e-OLX6-t, 

6(5/WGJ-d g-OLXJ7"C) 

8-0 LxO-q 

,,(6/w9j-dg-OLX6'6) 

,_-OLXS-t7 

I-OLXS;-c 

8-0 vs. L 

0ý-Ovvq 

6-0 V3- L 

oý-OLX6-L 

6-OLXL-6 

0ý-Ove-q 

8-0 vo-s 

6-0 Vt'-Z 

6-0 l`XO-9 

0ý-OLXL-z 

S-OLX8-z

-0 L X17- J7 

S-OLxt,*t) 

I-OLXZ-L 

6(6/W9J-dg-Ovt"q) 

I-OLXZ-L 

JB/wej-d L-0 LXO'Z) 

z-OLXS-6 

6-OLXZ-6 

Oj-OLXO-q 

Zý-OLXJ7-jý 

Ot-OLXZ-3 

ýt-OV6'C 

ýý-Ovvs 

zý-OLXL-L 

0ý-OLXL-L 

0ý-OLXL-L 

OL-OLXL-17 

e0V8-6 

6-0 L xs- L

8-OLxtl-6 

,(6/wei-d 

r,(6/w9j-d,,-OLxv0 

g-OLXZ-q 

.(6/wei-d -olxt,*/-) 

g-OLXE.c 

ýý-Ovvz 

L-0 V9-9 

6-OV8-t 

9-ovvc 

t -0 vv L 

9-0 vq. L 

9-0 vz- L 

8-0 Lxo-v 

8-Ovq-Z 

L-0 V6-9 

ZL-Ove-z

wvItI9 

,(Pajelclep) n 

,(Iejnzeu) n 

p(pinjeu) qL 

Odoýý 

I-L90z+ IEIWOLý 

UVO 

n3,,, 

wd,,,, 

eewLcý + SOICI 

eGS61 

16Ft 

15VW60t + P060L 

q)jgot + n8go, 

3166 

A06 + j S 06 

1>198 

'Neg 

0009 

9=129 

OS911 

loge 

014

6-0 LX9.9 

g-OLX17.9 

9-0 vvq

ot-mxv L 

6-OLX6-t, 

OL-OLXV L 

6-OLXZ.L 

0ý-Ovq-q 

m-OV8"C 

ý-O [xz- L 

,ejnsodx3 U01

Zý-mxvc 

ot-ovz-z 

jýj-mxq-z 

Lt-OLX6Z 

LL-OLXZ-z 

zt-OLXvz 

n-OLXZ-6 

isaiSul ejnsodxa uoijelinjul ains 

q(Q?1/WGJ-Uosjed) 

jualeAinb3 GSO(3 OAIIOG113 OA1130111103
Footnotes to Table A.2.2

a Models and parameter values are presented in Appendix A.2.3.1. Except as noted for thorium 

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from the 

disposal of a unit activity of 1 pCi (37 kBq) of radionuclides per year in all landfills. DSRs for all 

exposure pathways should be reduced by a factor of 0.8 if a unit activity of 1 /Ci (37 kBq) of 

radionuclides per year is assumed to be disposed in landfills plus incinerators (see 

Appendix A.2.3.1.5). 

b 1 person-rem/gCi = 0.27 person-Sv/Bq. 

c Depending on the physical form of the particular items of concern, DSRs for the exposure 

pathway could be reduced whenever radioactive materials should be significantly less 

dispersible than loose materials in trash (see Appendix A.2.3.1.5). 

d Calculated DSRs take into account the contributions from the decay products of 23•Th, which 

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and 

activity abundances of 232Th and • 8Th in natural thorium are given in Table 3.1.1. e Value in units of dose per unit mass is based on the assumed activity abundances of 232Th 

and "

8Th and the specific activity of 232Th given in Table 3.1.3 (1 p-Ci = 4.59 g). 

'Calculated DSRs take into account the contributions from the short-lived decay products of 21U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); 

assumed mass and activity abundances of 2_U, 235U, and 2 4U in natural uranium are given in 

Table 3.1.1. 

9 Value in units of dose per unit mass is based on the assumed activity abundances of 2

3U, 235

U, and 23U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 4Ci = 1.46 g). "h Calculated DSRs take into account contributions from the short-lived decay products of 2nU 

and 235

U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); 

assumed mass and activity abundances of 218U, 235U, and 234U in depleted uranium are given in 

Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 218

U, 2

'U, and 214U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 uCi = 2.7 g).

A.2-35
Table A.2.3 Individual Dose-to-Source Ratios for Exposure to Workers 

at Municipal Landf ills a

Radionuclide

36CI 

46SC 

55Fe 

6000 

63

Ni 

85Kr 

90

Sr + 90

Y 

99TC 

1

06Ru + 1

06Rh 

109Cd + l09m Ag 

1291 

137

cs + l

37mBa 

204 TI 

2

10m Bi +206 T1 

Th (natural)' 

U (natural)' 

U (depleted)' 

241

Am

External Expo~

2.Oxl Q-1 

2.5A1 0`0

Individual Effective Dose Equivalent 

(rernh/p4C)b 

sure Inhalation Exposure' Ingesi 

3.Oxl Q18 

6.4x1 0-17 

6.7x1 0-16 

9.1 x10-116 

8.3x1 0-17 

6.7x1 0-15 

9.5x1 0-17

Jion Exposurec 

7.Oxl Q-16 

2.3x1 0-14 

3.3xl 0-14 

7.Oxl 0-14 

6.7x10-15 

3.Oxl Q13 

6.3x 10-'5

1.X1 x -13

2.Oxl Q15 

2.0x10-" 

4.2x1 013 

2.Oxl 0-1 

3.1x1011 

5.7x10-" 

7.9x 0-16 

6.4x1 0-14 

2.2x10-" 

8.2x10>-16 

1 .3x 0`'0 

(2.8x10>' 1 rem/g)e 

1 .3x10>-12 

(8.9X10-13 remlg)g 

2.1 Xl0<' 2 

(7.6x 10 1-3 remlg)' 

6.9x1 0-13

4.Oxl 0-14 

3.1 xl Ql 

l.5x1014 

3.5x10-15 

5.3x1 015 

2.4x1 Q16 

9.8x1 Q16 

1 .2x 0-15 

6.8x10 15 

7.4x10-17 

2.6x1 Ql4 

2.9x1 Q13 

2.9x1 Qll 

(6.3x1 Q-12 remlg)e 

3.9x1 Q-12 

(2.7x1 Q12 remlg)g 

3.7x1 Q-12 

(1 .4x1 Q-12 rem/g)y 

1 .4x10'11

1 .7x1 Ql2 

1 .6x10'-4 

3.Oxl Q1,3 

1.4x10 13 

3.Oxl Q12 

3.7x1 0-14 

5.5x1 011 

1.2x 0-14 

7.1 X10' 4 

3.7x1 Ql4 

2.1x10' 1

1 

2.7x 0-11 

(6.Oxl 0-12 rem/g)e 

3.Oxl Q12 

(2.1 xl 10-2 remlg)g 

3.Oxl Ql2 

(1.1 X1 0- 2 remlg)i

4.0x1011

See following page for footnotes.

A.2-36

3

H
Footnotes to Table A.2.3

a Models and parameter values are presented in Appendix A.2.3.2. Except as noted for thorium 

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from the 

disposal of a unit activity of 1 gCi (37 kBq) of radionuclides per year in all landfills. DSRs for all 

exposure pathways should be reduced by a factor of 0.8 if a unit activity of 1 UCi (37 kBq) of 

radionuclides per year is assumed to be disposed in landfills plus incinerators (see 

Appendix A.2.3.1.5). 

b 1 rem//MCi = 0.27 Sv/Bq. 

' Depending on the physical form of the particular items of concern, DSRs for the exposure 

pathway could be reduced whenever radioactive materials should be significantly less 

dispersible than loose materials in trash (see Appendix A.2.3.1.5). 

d Calculated DSRs take into account contributions from decay products of 232Th, which are 

assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and activity 

abundances of 232Th and "

8Th in natural thorium are given in Table 3.1.1. e Value in units of dose per unit mass is based on the assumed activity abundances of 232Th 

and 28Th and the specific activity of 232

Th given in Table 3.1.3 (1 IpCi = 4.59 g). 

1 Calculated DSRs take into account contributions from short-lived decay products of 23"U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed 

mass and activity abundances of 238

U, 235

U, and 23U in natural uranium are given in Table 3.1.1. 

g Value in units of dose per unit mass is based on the assumed activity abundances of 238U, 235U, and 2

3U and the specific activities of uranium isotopes given in Table 3.1.2 

(1 /Ci = 1.46 g). 

" Calculated DSRs take into account contributions from the short-lived decay products of 21U 

and 231U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); 

assumed mass and activity abundances of 21U, 235U, and 234U in depleted uranium are given in 

Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 238U, 235

U, and 23U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 gCi = 2.7 g).

A.2-37
Table A.2.4 Collective Dose-to-Source Ratios for Exposure to Workers 

at Municipal Landfills a

Radionuclide

14C 

36CI 

46

sc 

55 Fe 

60C0 

63 Ni 

85 Kr 

90

S r + 90

Y 

99

Tc 

16u+ 106 Rh 

'

09Cd + 109

m Ag 

1291 

1

33Ba 

17S+ 1

3 7

m Ba' 

1

52 Eu 

204 TI 

2 1Om Bj ±20'TI 

210Po 

Th (natural)' 

U (natural)' 

U (depleted)'

External Expo

3.5A 0-6 

4.5x 0-6 

2.Oxl 0-'

3.5x 10-" 

3.5x1 0-7 

7.3xl 09 

3.6x1 0-9 

5.4x1 0

9.9X1 0-7 

1.4x1011 

1 .9X1 -6 

1.1 Xl -9 

3.8x1 0-7 

1 .4x1011 

2.2x1 0-6 

(4.8x1 0-7 p-rem/g)e 

2.3x10 8 

(1 .6x1 -8 p-rem/g)g 

3.6x1 Q8 

(1 .3x1 0-8 p-remlg)' 

1 .2x1 O-"

7.Oxl0-10 

5.5x1 0-13 

2.6x1 10-' 

6.2x1 01¶ 

9.3x-1 011 

4.2x1 0-12 

1 .7x10-" 

2.lxlO'1 

1.2x10-10 

1.3x 0-12 

4.5x1 Q0` 

5.1 x10-9 

5.1 X1l

(1.1X10-7 p-rem/g)e 

6.8x10-8 

(4.6x1 Q8 p-rem/g)g 

6.5x1 08 

(2.4x-10-8 p-rem/g)' 

2.4x1 0-7

ion Exposurec 

1 .2x101 ' 

4.Oxl Q0-' 

5.8x 1 Q10 

1.2x1 0

1 .2x10Q10 

5.2x1 09 

1.1X10-10

2.9x 10-' 

2.8x10-10 

5.3x100-1 

2.5x1 Q-9 

5.3x1 0-8 

6.5x` 10-1 

9.6x1 0' 

2.Oxl 0-10 

1.2x 0-9 

6.5x-10-10 

1.8x100

3.7x 10-' 

4.8x 0-7 

(1 .0X10 7 p-rem/g)e 

5.3x1 0-' 

(3.6x1 Q8 p-remlg)g 

5.2x1 Q8 

(1 .9x1 0-8 p-remlg)i 

7.Oxl 0-7

See following page for footnotes.

A.2-38

3

H

Collective Effective Dose Equivalent 

(person-rernlsCi)b 

'sure Inhalation Exposurec Ingest 

5.2A1 -14 

1.1 X10-12 

1 .2x1011 

1 .6x1 Q-1 

1 .4x1 0-12 

1 .2x-10-10 

1 .7x1 0-12
Footnotes to Table A.2.4

a Models and parameter values are presented in Appendix A.2.3.2. Except as noted for thorium 

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from the 

disposal of a unit activity of 1 A.Ci (37 kBq) of radionuclides per year in all landfills. DSRs for all 

exposure pathways should be reduced by a factor of 0.8 if a unit activity of 1 /,Ci (37 kBq) of 

radionuclides per year is assumed to be disposed in landfills plus incinerators (see 

Appendix A.2.3.1.5). 

b 1 person-rem//2Ci = 0.27 person-Sv/Bq. 

I Depending on the physical form of particular items of concern, DSRs for the exposure 

pathway could be reduced whenever radioactive materials should be significantly less 

dispersible than loose materials in trash (see Appendix A.2.3.1.5). 

d Calculated DSRs take into account the contributions from the decay products of 232

Th, which 

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and 

activity abundances of 232Th and •

8Th in natural thorium are given in Table 3.1.1. 

e Value in units of dose per unit mass is based on the assumed activity abundances of 232

Th 

and "

8Th and specific activity of 232Th given in Table 3.1.3 (1 uCi = 4.59 g). 'Calculated DSRs take into account the contributions from the short-lived decay products of 238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed mass and activity abundances of 23U, 235U, and 234U in natural uranium are given in 

Table 3.1.1. 

g Value in units of dose per unit mass is based on the assumed activity abundances of 238U, 2

1U, and 2

'U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 1,Ci = 1.46 g). 

h Calculated DSRs take into account the contributions from the short-lived decay products of 21 U and 21U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed mass and activity abundances of 238U, 21U, and 2

'U in depleted uranium are given in 

Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 238U, 235U, and 214U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 1/ Ci = 2.7 g).

A.2-39
Table A.2.5 Individual and Collective Dose-to-Source Ratios for Exposure to Off-Site 

Residents at Municipal Landfills Due to Airborne Releases During 

Landfill Operations a

Radionuclide 

3

H 

14C 

36CI 

4 6

Sc 

55

Fe 

6c0o 

63

Ni 

85

Kr 

90

Sr + 9

0

Y 

99Tc 

'

0 6Ru + '

06

Rh 

109Cd + l1 9mAg 

1291 

1

3Ba 

1370s + 1

3 7 MtBa 

147pm 

1

5 2Eu 

204TI 

21OmBi + 2°6TI 

21 0

Po 

Th (natural)' 

U (natural)' 

U (depleted)i

Individual Effective Dose Equivalentb,c 

(renm/gCi)d 

1.1x10-

17 

1.7x10-16 

9.1x10-14 

1.6x10-15 

4.4x1 0-17 

3.0x1 0-14 

4.4x1 0-17 

2.9x1 0-18 

2.2x10-15 

1.9x10-'5 

3.7x1 0-15 

1.2x10-15 

4.3x1 0-14 

9.2x1 0-15 

4.3xl 0-'5 

2.2x1 0-16 

3.0x1 0-14 

3.4x10 16 

6.0x10-15 e 

9.1x 10-14 

1.5x10-

12 

(3.3x10- 3 rem/g)g 

6.0xl 0-13 

(4.1xl0 13 rem/g)' 

5.7x10-13 

(2.1xl 0-13 rem/g)k 

2.7x10-12

See following page for footnotes.

A.2-40

Collective Effective Dose 

Equivalentb', 

(person-rem/UCi)d 

3.3x1 0-12 

3.0x10-l1 

1.5x1 0-7 

11.6x10-9 

5.5x10l" 

3.1x10-8 

3.5x10-11 

4.5x1 0- 12 

1.4x10-9 

1.8x10-9 

2.5x1 0-9 

7.7x10-o1 

6.0x1 0-9 

9.4x10-9 

5.2x1 0-9 

1.4x10-10 

3.1 x1 0

5.3x1 0-10 

3.7x10o-e 

5.4x10-8 

9.1x10-7 

(2.0xl 0-7 p-rem/g)g 

3.5x1 0-7 

(2.4xl 0-7 p-rem/g)' 

3.3x10-7 

(1.2x10-7 p-rem/g)k 

1.6x10-6
Footnotes to Table A.2.5

a Models and parameter values are presented in Appendix A.2.3.3. Except as noted for thorium 

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from the 

disposal of a unit activity of 1 ,aCi (37 kBq) of radionuclides per year in all landfills. 

b DSRs should be reduced by a factor of 0.8 if a unit activity of 1 MCi (37 kBq) of radionuclides 

per year is assumed to be disposed in landfills plus incinerators (see Appendix A.2.3.1.5). 

c Depending on the physical form of the particular items of concern, DSRs could be reduced 

whenever radioactive materials would be significantly less dispersible than loose materials in 

trash (see Appendix A.2.3.1.5). 

1 rem//Ci = 0.27 Sv/Bq; 1 person-rem/mCi = 0.27 person-Sv/Bq. 

e Value estimated based on the calculated DSR for Th (natural), which results primarily from 

inhalation exposure, and the assumption that the dose from 2lomBi also results primarily from 

inhalation exposure. 

'Calculated DSRs take into account the contributions from the decay products of 232Th, which 

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and 

activity abundances of 232Th and 228

Th in natural thorium are given in Table 3.1.1. 

g Value in units of dose per unit mass is based on the assumed activity abundances of 232Th 

and 228

Th and the specific activity of 232

Th given in Table 3.1.3 (1 1/ Ci = 4.59 g). h Calculated DSRs take into account the contributions from the short-lived decay products of 238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed mass and activity abundances of 238

U, 235

U, and 2-U in natural uranium are given in 

Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 238

2 U, 35U, and 23U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 ACi = 1.46 g). 

J 

2

Calculated DSRs take into account the contributions from the short-lived decay products of 

1U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed mass and activity abundances of 238U, 235U, and 23U in depleted uranium are given in 

Table 3.1.1. 

k Value in units of dose per unit mass is based on the assumed activity abundances of 238U, 

235

U, and 21U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 gCi = 2.7 g).

A.2-41
Table A.2.6 Solid and Solution Distribution Coefficients (Kd) for Elements a 

Kd Kd 

Element (mLug) Element (mug) 

H 0 I 1.0 

C 6.7 Ba 5.2x1 0' 

Cl 1.7 Cs 2.7x 102 

Sc 3.1x10 2 Pm 2.4x102 

Fe 1.6x10 2 Eu 2.4x10 2 

Co 6.0x10 1 TI 3.9x10 2 

Ni 4.0x10 2 Bi 1.2x10 2 

Sr 1.5x10, Po 1.5x10 2 

Tc 1.0x10- 1 Th 3.2x10 3 

Ru 5.5x10 1 U 1.5x10' 

Cd 4.Ox101 Am 1.9x10 3 

a Values obtained from Table 6.7 of NUREG/CR-5512.

A.2-42
Table A.2.7 Individual and Collective Dose-to-Source Ratios for Exposure to Off-Site 

Residents at Municipal Landfills Due to Releases to Groundwater a 

Individual Effective Collective Effective 

Timeb Dose Equivalentc'd Dose Equivalentcd,e 

Radionuclide (yr) (rem/gCi)l (person-rem/gCi)l

3

H 

140

1.0x10, 

3.2xl 02 

8.9x1 01 36Ci 

46

Sc 

55

Fe 

60Co 

63 Ni 

90Sr + 90Y 

99Tc

7.1x10 2 

1.5x101

1

0 6Ru + 1°6Rh 

1

09Cd + lo0mAg 

1291 

1

33

Ba 

1370s + 1

37mBa 

147pm 

I

52 Eu 

204

T1 

210mBi + 206

TI 

210po 

Th (natural)9 

U (natural)' 

U (depleted)k

5.7xl 01

5.6xl 03 

1.5x1 05 

7.1x10

2 

7.1x10 2

1.2x10-1' 

4.3x10-15 

2.3x1 0-14

3.3x1 0-21 

6.8x1 0-14

3.3x1 0-12

1.0x10-9 

6.5x10-8 

4.9x10-'

4.5x10-11 

1.6x10-6

7.3x1 0-5

1.2x1 0-14

1.1xlO-14 

(2.4xl 0- 15 rem/g)h 

2.7x10-13 (1.8x1 0-13 rein/g)i 

2.6x10-13 

(9.6xl 0-14 rem/g)'

1.7x10-6 

(1.2x1 0-6 p-rem/g)i 

1.7x10-6 

(6.2xl 0-7 p-rem/g)'

See following page for footnotes.

A.2-43
Footnotes to Table A.2.7

a Models and parameter values are presented in Appendix A.2.3.4. Except as noted for thorium 

and uranium, dose-to-source ratios (DSRs) give annual individual effective dose equivalents (EDEs) from the disposal of a unit activity of 1 uCi (37 kBq) of radionuclides per year in all 

landfills over 30-year operating lifetime and collective EDEs for 1,000 years after facility closure 

from disposal of a unit activity of 1 yCi (37 kBq) of radionuclides in all landfills during the last 

year of operations only. 

b Travel time of radionuclide in groundwater from landfill to off-site municipal well. If no entry is 

given, the travel time is much greater than the half-life of the radionuclide, and the resulting 

doses are essentially zero (0). 

C DSRs should be reduced by a factor of 0.8 if a unit activity of 1 YCi (37 kBq) of radionuclides 

per year is assumed to be disposed in landfills plus incinerators (see Appendix A.2.3.1.5). 

d Depending on the physical form of the particular items of concern, DSRs could be reduced 

whenever radioactive materials should be significantly less accessible to infiltrating water than 

loose materials in waste (see Appendix A.2.3.4.3). e Collective dose is calculated only if the travel time of the radionuclide in groundwater is less 

than 1,000 years. 

f 1 rem/pCi = 0.27 SvIBq; 1 person-rem/mCi = 0.27 person-Sv/Bq. 

g Calculated DSRs take into account the contributions from the decay products of 232Th, which 

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and 

activity abundances of 232

Th and "

8Th in natural thorium are given in Table 3.1.1. 

" Value in units of dose per unit mass is based on the assumed activity abundances of 232Th 

and 28Th and the specific activity of 232Th given in Table 3.1.3 (1 MCi = 4.59 g). 'Calculated DSRs take into account the contributions from the short-lived decay products of 238 U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); 

assumed mass and activity abundances of 238U, 235U, and 234U in natural uranium are given in 

Table 3.1.1. 

J Value in units of dose per unit mass is based on the assumed activity abundances of 23U, 21U, and 2

`U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 MCi = 1.46 g). 

k Calculated DSRs take into account the contributions from the short-lived decay products of 238

U and 235

U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); 

assumed mass and activity abundances of 238U, 235

U, and 234U in depleted uranium are given in 

Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 2 8U, 235U, and 2

'U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 uCi = 2.7 g).

A.2-44
Table A.2.8 Individual Dose-to-Source Ratios for Exposure to Future On-Site 

Residents at Municipal Landfills a

Radionuclide

14C 

36CI 

468cd 

55

Fe 

60Co 

63

Ni 

85Kr 

9OSr + 90y 

99

Tc 

106 Ru + 106 Rh 

109Cd + l° 9

mAg 

1291 

1&3Ba 

137

Cs + 137mBa 

147 Pm 

1

5 2 Eu 

2O4TI 

21OmBi + 206

TI 

210pod 

Th (natural)e 

U (natural)h 

U (depleted)j

External Expos

4.1x10- 2 

2.4x1 0-14

6.5x1 0-15 

3.7xl 0-21 

6.7xl 0-21 

6.7xl 0-13 

6.6x10-12 

6.8x10-11 

1.2x10- 19 

4.1x10-1' 

1.6x10-1 6 

7.2x10- 11 

4.2x10-10 

(9.2x1 0-11 rem/g)f 

4.3x1 0- 12 

(2.9x1 0-12 rem/g)' 

6.8x1 0- 12 

(2.5xl 0-12 rem/g)k 

2.lxl 0-12

Individual Effective Dose Equivalent 

(rem/lCi)b 

sure Inhalation Exposurec Ingest 

4.2xl 0-20 

1.5x1 0-` 

1.6x10-16 

1.2x10-21 

7.8x10-'8 

1.7x10-17

3.3x10-15 

9.8x10-

15 

1.9x1 0-25 

4.1 x1 0-24 

1.3x1 0-15 

3.7x1 0-18 

8.6x1 0-17 

1.3x1 0-20 

1.8x10-16 

1.3x1 0-20 

5.6x1 0-14 

1.0x10-9 9 

(2.3x10-° rem/g)f 

9.3x1 0-13 

(6.4x10-13 remi/g)' 

8.9x1 0-13 

(3.3x10-13 rem/g)k 

3.1x10-12

tion Exposure' 

1.2x10-18 

.4x1 0-16 

3.4xl 0-'6 

.1x10-2

0 

2.7x1 0-17 

3.9xl 0-17

1.1x10-14 

3.1x10-16 

3.1x10-

25 

1.3x 0-23 

5.8x10-14 

4.6x1 0-17 

3.8x1 0-15 

1.0xl 0-20 

1.5x1 0-16 

5.2x1 0-19 

2.0x1 0-14 

2.6x10-13 

(5.7x10-14 rem/g)f 

5.8x10-14 

(4.Oxl 0-14 rem/g)' 

5.7x10-14 

(2.1 xl0-14 rem/g)k 

7.1x1 0-13

See following page for footnotes.

A.2-45

3H
Footnotes to Table A.2.8

a Models and parameter values are presented in Appendix A.2.3.5. Dose-to-source ratios 

(DSRs) give annual individual effective dose equivalents from the disposal of a unit activity of 

1 yCi (37 kBq) of radionuclides per year in all landfills over 30-year operating lifetime. DSRs 

should be reduced by a factor of 0.8 if a unit activity of 1 kiCi (37 kBq) of radionuclides per year is assumed to be disposed in landfills plus incinerators (see Appendix A.2.3.1.5). b 1 rem/pCi = 0.27 Sv/Bq. 

Depending on the physical form of the particular items of concern, DSRs for the exposure 

pathway could be reduced whenever radioactive materials should be significantly less 

dispersible than loose materials in trash (see Appendix A.2.3.5.5). 

d Time after facility closure for first occurrence of on-site residence is much greater than the 

half-life of the radionuclide, and the resulting doses are essentially zero (0). 

e Calculated DSRs take into account the contributions from the decay products of 232Th, which 

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and 

activity abundances of 232Th and "

8Th in natural thorium are given in Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 232Th and 

"•8Th and the specific activity of 23"'Th given in Table 3.1.3 (1 uCi = 4.59 g). 

g Value is the dose from exposure to 220

Rn and its short-lived decay products during indoor 

residence (see Appendix A.2.3.5.2). Contribution from inhalation exposure to 232Th and its 

other decay products is 3.5x10-12 rem/yCi and, thus, is negligible by comparison. 

h 

2

Calculated DSRs take into account the contributions from the short-lived decay products of 

1 U and 21U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed mass and activity abundances of 238U, 235U, and 2

"U in natural uranium are given in 

Table 3.1.1. 

SValue in units of dose per unit mass is based on the assumed activity abundances of 2

38U, 235

U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 uCi = 1.46 g). 

J Calculated DSRs take into account the contributions from the short-lived decay products of 238U, and 235

U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed mass and activity abundances of 238U, 235

U, and 234U in depleted uranium are given in 

Table 3.1.1. 

k Value in units of dose per unit mass is based on the assumed activity abundances of 2

2 38U, 35

U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 /Ci = 2.7 g).

A.2-46
Table A.2.9 Collective Dose-to-Source Ratios for Exposure to Future On-Site 

Residents at Municipal Landfills a

Radionuclide

14C 

36CI 

46Scd 

55

Fe 

60Co 

63Ni 

85

Kr 

90Sr + 90y 

106Ru + 106Rh 

109Cd + lO9mAg 

1291 

133 Ba 

"

137Cs + 137mBa 

14 7 Pm 

15 2Eu 

204TI 

21OmBi + 206TI 

210p o d 

Th (natural)e 

U (natural)h 

U (depleted)' 

241

Am

External Expo

1.5x10-7

Collective Effective Dose Equivalent 

(person-rem//gCi)b 

'sure Inhalation Exposurec Ingest 

1.8xI0-'5 

1.7x10-11 

1.9x10-10 

4.1x10- 7 

2.8x1 0-13 

3.2x1 0-12

tion Exposurec 

5.1x10-

14 

4.8x10-1o 

7.4x10-10 

1.5x10-15 

9.7x10-13 

1.7x10-11

9.7x10-10

7.6x10-9 

1.3x10-16 

2.3x1 0-16 

7.9x10-7 

2.7x1 0-7 

4.8xl 0-6 

4.2x10-' 5 

1.8x10-6 

5.5x10-12 

8.4xl 0'

4.9x1 0-4 

(1.1x10- 4 p-rem/g)f 

4.5x1 0-6 

(3.1x1 0-6 p-rem/g)' 

6.9x1 0-6 

(2.6x10-6 p-rem/g)k 

1.3xl 0-6

2.3x10-'o 

1.1x10-8 

6.7xl 0-21 

1.4x10-19 

1.5x10-9 

1.5x10-' 3 

6.0x1 0-12 

4.6x1 0-16 

8.1 x10-12 

4.6x1 0-16 

6.6x10-8 

1.2x10-3 g 

(2.6xl 0-4 p-rem/g)f 

1.1x10-6 

(7.4xl 0-7 p-rem/g)' 

1.0x10-6 

(3.8x10-` p-rem/g)" 

1.8x10-6

7.5x10-Q1 

3.6x10-10 

1.1x10-2

0 

4.7x10-' 9 

6.8x10 8 

1.9x10-

12 

2.7x10-1' 

3.5x10-16 

6.7x10-12 

1.8x1 0-14 

2.3x1 0-8 

6.1x10-7 (1.3Xl10-7 p-rem/g)' 

6.7x10-8 

(4.6xl 0-8 p-rem/g)' 

6.6x1 0-8 

(2.4x10-8 p-rem/g)k 

4.2x1 0-7

See following page for footnotes.

A.2-47
Footnotes to Table A.2.9

a Models and parameter values are presented in Appendix A.2.3.5. Except as noted for thorium 

and uranium, dose-to-source ratios (DSRs) give collective effective dose equivalents for 

1,000 years after loss of institutional controls over disposal facilities from disposal of a unit 

activity of 1 /.Ci (37 kBq) of radionuclides per year in all landfills during the last year of 

operations only. DSRs should be reduced by a factor of 0.8 if a unit activity of 1 /2Ci (37 kBq) of 

radionuclides per year is assumed to be disposed in landfills plus incinerators (see 

Appendix A.2.3.1.5). 

b 1 person-rem/MuCi = 0.27 person-Sv/Bq. 

c Depending on the physical form of the particular items of concern, DSRs for the exposure 

pathway could be reduced whenever radioactive materials should be significantly less 

dispersible than loose materials in trash (see Appendix A.2.3.5.5). 

I Time after facility closure for first occurrence of on-site residence is much greater than the 

half-life of the radionuclide, and resulting doses are essentially zero (0). e Calculated DSRs take into account the contributions from the decay products of 232

Th, which 

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and 

activity abundances of 232Th and 2

8Th in natural thorium are given in Table 3.1.1. 

f Value in units of dose per unit mass is based on the assumed activity abundances of 232

2 Th and 28

Th and the specific activity of 232

Th given in Table 3.1.3 (1 1.Ci = 4.59 g). 

g Value is dose from exposure to 22°Rn and its short-lived decay products during indoor 

residence (see Appendix A.2.3.5.2). Contribution from inhalation exposure to 232

Th and its 

other decay products is 8.2x10- 6 person-rem/gCi and, thus, is negligible by comparison. 

h Calculated DSRs take into account the contributions from the short-lived decay products of 232U and 235

U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed mass and activity abundances of 2

3U, 235U, and 2

3U in natural uranium are given in 

Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 2 1U, 

"235

U, and 2

'U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 /_Ci = 1.46 g). 

J Calculated DSRs take into account the contributions from the short-lived decay products of 2 1U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); 

assumed mass and activity abundances of 21U, 235U, and 2

34U in depleted uranium are given in 

Table 3.1.1. 

k Value in units of dose per unit mass is based on the assumed activity abundances of 238U, 231U, and 2

'U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 MCi = 2.7 g).

A.2-48
Table A.2.10 Individual Dose-to-Source Ratios for Exposure to Waste 

Collectors at Municipal Incinerators a

Radionuclide

14C 

36CI 

46Sc 

55Fe 

60

Co 

63 Ni 

85Kr 

' 0

Sr +90 

99TC 

19d+ 109m Ag 

1291 

1

33 Ba 

13

1CS + 1

37

m Ba 

1

52Eu 

204

Ti 

2

1lOmBi + 206 TI 

Th (natural)d 

U (natural)' 

U (depleted)h 

241

Am

External Expo

3.6x1 0-8 

4.3x 0-8

Individual Effective Dose Equivalent 

(reM/4Ci)b 

sure Inhalation Exposuree Ingest 

6.1 Xl0 16 

1 .4x1 0-14 

1 .4x1 Q13 

1 .9x1 0-13 

1.8x 0-14 

1 .4x1 Q-12 

2.Oxl 0-14

ion Exposurec 

7.8x1 0-14 

2.5A1 -12 

3.7x1 Q12 

7.8x1 Q-12 

7.4x1 O-' 

3.3x101 ' 

7.Oxl Q-1

4.4x1 Q11

1 .5x1Q"-4 

3.9x 0-9 

1.8x10-10 

2.7x1 a¶O 

8.X1 0-9 

1.1 Xl0 

9.3x1 0-14 

2.1 x10-8 

3.2x 0-11 

5.8x1 09 

1.6x10'3 

2.2x1 Q-8 

(4.8x10-9 remlg)e 

3.5X10-10 

(2.3x-1 -10 rem/g)g 

5.2x1 Q10 

6.3x10 'o

8.6x1 0-12 

6.6x1 0-15 

3.1 X1 Q-12 

7.5x1 -13 

1.1X10-12 

5.1 Xl10 14 

2.1 X10' 3 

2.6x 10-'3 

1 .4x1 0-12 

1 .6x1 0-14 

5.3x1 0-12 

6.2x1 0-" 

6.2x1 0-9 

(1.4x109' remlg)e 

8.2X10 1 0O 

(5.6x10 10 rem/g)g 

7.8x-1 Q>1 

(2.9x-10`1 remlg)' 

2.9x10 9

1 .9x10-10 

1 .8X10-12 

3.3x 10- " 

1 .6x"10-1' 

3.4x1 -lO 

4.1 X10' 2 

6.1 xl Qi 

1 .3X10-12 

7.9x1 0-12 

4.1 Xl0>' 2 

1.2X10`10 

2.3X1 0-9 

3.Oxl 09 

(6.6x1 O-10 remlg)e 

3.4x10-10 

(2.3x 10- 10 remlg)g 

3.3x"1 Q>1 

(1 .2x 10 10 re m/g)' 

4.4x-10-9

See following page for footnotes.

A.2-49

3

H
Footnotes to Table A.2.10

a Models and parameter values are presented in Appendix A.2.4.1. Except as noted for thorium 

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from 

disposal of a unit activity of 1 Ci (37 kBq) of radionuclides per year in all incinerators. DSRs for 

all exposure pathways should be reduced by a factor of 0.2 if a unit activity of 1 /Ci (37 kBq) of 

radionuclides per year is assumed to be disposed in landfills plus incinerators (see 

Appendix A.2.3.1.5 and A.2.4.1). b 1 rem/kzCi = 0.27 Sv/Bq. 

Depending on the physical form of the particular items of concern, DSRs for the exposure 

pathway could be reduced whenever radioactive materials should be significantly less 

dispersible than loose materials in trash (see Appendix A.2.3.1.5). 

d Calculated DSRs take into account the contributions from the decay products of 232

Th, which 

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and 

activity abundances of 232Th and "STh in natural thorium are given in Table 3.1.1. e Value in units of dose per unit mass is based on the assumed activity abundances of 232Th 

and 2 8

Th and the specific activity of 232Th given in Table 3.1.3 (1 /Ci = 4.59 g). f Calculated DSRs take into account the contributions from the short-lived decay products of 238

U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed mass and activity abundances of 238

U, 231U, and 2 •U in natural uranium are given in 

Table 3.1.1. 

g Value in units of dose per unit mass is based on the assumed activity abundances of 238U, 235

U, and 23U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 ýCi = 1.46 g). 

h Calculated DSRs take into account the contributions from the short-lived decay products of 238

U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed mass and activity abundances of 238U, 235U, and 2 1U in depleted uranium are given in 

Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 2-8U, 2 1U, and 21U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 gCi = 2.7 g).

A.2-50
Table A.2.1 1 Collective Dose-to-Source Ratios for Exposure to Waste 

Collectors at Municipal Incinerators a

Radionuclide

140 

36CI 

55 Fe 

60C0 

63 Ni 

85Kr 

90Sr + 90Y 

99TC 

1

06 Ru + 16R 

19d+ loomAg 

1291 

133 Ba 

'

370s + l

37

mBa 

1

52 Eu 

204

TI 

2

10m Bi + 206

TI 

Th (natural)d 

U (natural)f 

U (depleted)'

External Expo

5.4x1 0-6 

6.4x1 0-6 

6.6x1 0-9

2.3x1 0-12 

5.9x1 0-7 

2.8x1 0-" 

4.Oxl 0-" 

1.2x 0-6 

1.6x 0-6 

1.4x10>'1 

3.1 x1 0

4.8x 10-9 

8.6x1 0-7 

2.4x10' 11 

3.3x1 0-6 

(7.4x1 0-7 p..rem/g)e 

5.2x 0-8 

(3.4x1 0-" p-remfg)g 

7.8x1 0-8 

(2.8x1 -8 p-rem/g)i 

9.4x-1 0-

Collective Effective Dose Equivalent 

(person-reM1IpCi)b 

sure Inhalation Exposurec Ingest 

9.2x10'-4 1 

2.1 Xl102 

2.2x10'1 

2.9x10-" 1 

2.6x1 0-12 1 

2.2x10-10 4 

3.1 X10' 2 1

1.3x 0-9 

9.8X10-13 

4.7x1 0`0 

1 .1X10'10 

1 .7x-10-1'0 

7.7x1 Q-12 

3.1X10"1 

3-9X10-1" 

2.2x1 Q10) 

2.4x1 Q-12 

8.0X10>' 0 

9.2x 0-9 

9.3x10-7 

(2.OxlO07 p-rem/g)e 

1.2x1 0-7 

(8.4x1 08 p-rem/g)g 

1 .2x 0-7 

(4.4x1 0-8 p-rem/g)y 

4.4x1 0-7

ion Exposurec 

I.2x1 Q11 

~.8x1 0-'O 

~.5X1 0-10 

.2x1 09 

.1X10-10 

.9x1 Q-9 

.1 Xl QlO

2.8x1 0-8 

2.7x 10- 10 

5.Qxl 09 

2.4x 10Q

5.Oxl Q8 

6.2x1 Q10` 

9.1 X10-9 

1 .9X10-10 

1 .2x 0-9 

6.1 xl -10 

1.8x10-8 

3.5x1 0-7 

4.5x1 0-7 

(9.9x1 08 p-reM/g)e 

5.OXl 0-8 

(3.4x1 0-' p-remlg)g 

4.9xl 0-8 

(1.8x108

' p-rem/g)' 

6.6x 10-'

See following page for footnotes.

A.2-51
Footnotes to Table A.2.1 1

a Models and parameter values are presented in Appendix A.2.4.1. Except as noted for thorium 

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from 

disposal of a unit activity of 1 pzCi (37 kBq) of radionuclides per year in all incinerators. DSRs for 

all exposure pathways should be reduced by a factor of 0.2 if a unit activity of 1 ptCi (37 kBq) of 

radionuclides per year is assumed to be disposed in landfills plus incinerators (see 

Appendix A.2.3.1.5 and A.2.4.1). b 1 person-rem/yCi = 0.27 person-Sv/Bq. 

Depending on the physical form of the particular items of concern, DSRs for the exposure 

pathway could be reduced whenever radioactive materials should be significantly less 

dispersible than loose materials in trash (see Appendix A.2.3.1.5). 

d Calculated DSRs take into account the contributions from the decay products of 232Th, which 

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and 

activity abundances of 232Th and "

8Th in natural thorium are given in Table 3.1.1. e Value in units of dose per unit mass is based on the assumed activity abundances of 232

Th 

and "

6Th and the specific activity of 232Th given in Table 3.1.3 (1 UCi = 4.59 g). f Calculated DSRs take into account the contributions from the short-lived decay products of 2

`U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed mass and activity abundances of 238

U, 235U, and 21U in natural uranium are given in 

Table 3.1.1. 

g Value in units of dose per unit mass is based on the assumed activity abundances of 238

U, 235

U, and 2

34U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 pCi = 1.46 g). 

h Calculated DSRs take into account the contributions from the short-lived decay products of 238

U and 235

U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed mass and activity abundances of 2

38U, 235U, and 234U in depleted uranium are given in 

Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 238U, 21U, and 2 14U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 MCi = 2.7 g).

A.2-52
Table A.2.12 Individual Dose-to-Source Ratios for Exposure to Workers 

at Municipal Incinerators a

Radionuclide

140 

36CI 

46s0 

55 Fe 

60

co 

63Ni 

85Kr 

`0Sr +9Y 

99

Tc 

1

06 Ru + 16R 

1

09 0d + lo9mAg 

1291 

133 Ba 

17S+ l

37

mBa 

152 Eu 

204

TI 

21lOm B + 206

TI 

210Po 

Th (natural)' 

U (natural)' 

U (depleted)' 

24

'Am-241

External Expo~

3.9x10"1 

4.9x1 Q11 

2.2x10'4

3.8x 10 "l 

3.9x10'-2 

8.1 x1014 

3.9x1 O-' 

6.Oxl Q-12 

1.1X10-', 

2.1xl0>" 

1 .2x1014 

4.2x1 0-12 

1.6x10'-6 

2.5x101' 

(5.4x1 0-12 rem/g)e 

2.5x1 Q13 

(1 .7x10'-3 remlg)g 

4.Oxl Q-1 

(1.5X10-13 rem/gyi 

1 .3x1 O-'

Individual Effective Dose Equivalent 

(reM/,UCi)b 

sure Inhalation Exposurec Ingest 

1.7x10-'6 

3.7x1 0-17 

3.9x 0-16 1 

5.3x1 0-16 

4.8x1 0-17 

3.9x 0-15 9 

5.6x10'-7 1

2.3x1 0-14 

1 .8x10'7 

8.6x1 0` 5 

2.1x-10-15 

3.1xl0'5 

1 .4x1 0-16 

5.7x Q106 

7.Oxl 0-16 

4.OXl Q15 

4.3x1 0-17 

1 .5x10>-14 

1.7x10 13 

1 .7x-10-11 

(3.7x1 0-12 rem/g)e 

2.3x1 0-12 

(1 .5x1 0-12 remlg)g 

2.1 Xl102 

(7.9x1Q0-13 rem/g)' 

8.Oxl 0-12

ion Exposurec 

?.1 X10-17 

'.OX1 016 

~.1X10-15 

.QX 10 Q16 

~.OX1 Q15 

.9X10-16

5.1 xl 14 

4.9x1 0-16 

9.1x10-'5 

4.4x-10Q11 

9.2x1 0-14 

1 .7x1 0-14 

3.5x1 016 

2.2x1 cr' 5 

1.lxlO 15> 

3.2x1 0-14 

6.3x1 0-13 

8.3x1 0-13 

9.2x1 0-14 

(6.3x1 0-14 rem/g)g 

9.OX 0-14 

(3.3x1 0-14 rem/g)' 

1.2x10'-2

See following page for footnotes.

A.2-53
Footnotes to Table A.2.12

a Models and parameter values are presented in Appendix A.2.4.2. Except as noted for thorium 

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from 

disposal of a unit activity of 1 yCi (37 kBq) of radionuclides per year in all incinerators. DSRs for 

all exposure pathways should be reduced by a factor of 0.2 if a unit activity of 1 MCi (37 kBq) of 

radionuclides per year is assumed to be disposed in landfills plus incinerators (see 

Appendix A.2.3.1.5 and A.2.4.2.5). b 1 rem/puCi = 0.27 Sv/Bq. 

c Depending on the physical form of the particular items of concern, DSRs for exposure 

pathway could be reduced whenever radioactive materials should be significantly less 

dispersible than loose materials in trash (see Appendix A.2.3.1.5). 

d Calculated DSRs take into account the contributions from the decay products of 232Th, which 

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and 

activity abundances of 23 lTh and 228

Th in natural thorium are given in Table 3.1.1. e Value in units of dose per unit mass is based on the assumed activity abundances of 232Th 

and •8Th and the specific activity of 232Th given in Table 3.1.3 (1 gCi = 4.59 g). 

'Calculated DSRs take into account the contributions from the short-lived decay products of 23U and 235

U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); 

assumed mass and activity abundances of 238U, 235U, and 2 4U in natural uranium are given in 

Table 3.1.1. 

§ Value in units of dose per unit mass is based on the assumed activity abundances of 23U, 23U, and 21 U and specific activities of the uranium isotopes given in Table 3.1.2 

(1 yCi = 1.46 g). 

h Calculated DSRs take into account the contributions from the short-lived decay products of 238U and 235

U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); 

assumed mass and activity abundances of 2

3U, 235

U, and 234U in depleted uranium are given in 

Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 238U, "

2•U, and 2 14U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 uCi = 2.7 g).

A.2-54
Table A.2.13 Collective Dose-to-Source Ratios for Exposure to Workers 

at Municipal Incinerators a

Radionuclide

36CI 

46

SC 

55 Fe 

60 Co 

63Ni 

85Kr 

90Sr +90 

99

Tc 

106 Ru + 1

061h 

19d+ 109m Ag 

1291 

1

33 Ba 

1CS+ 137m Ba 

147 Pm 

1 52Eu 

2

10m Bi + 2T 

Th (natural)' 

U (natural)' 

U (depleted)'

External Expo~

1 .2x108I 

1.5x108'

Collective Effective Dose Equivalent 

(person-rem/gCI)' 

sure Inhalation Exposu rec Ingesti 

5.2x 0-16 E 

1.1X10-14 r 

1.2x 0-13 3 

1.6x 0-13 6 

1.4x1 0-14 6 

1.2x10'-2 2 

1.7x 0-14 5

on Exposurec 

A.x1 Q159 

~.1X1 0-13 

AxIl Q-13 

;.1X10-14 

.7x1 Q-12 

.8x1 0-14

6.5x1 0-12

1.2xl0'3 

1 .2A10-' 

2.4x1 Q11 

1.2x101 1 

1.8x 0-9 

3.3x1 0-9 

4.6x 0-14 

6.4x1 O-9 

3.7x1 0-12 

1 .3x 0-9 

4.8x1 0-l4 

7.4x1 0-9 

(1 .6x 0-9 p-rem/g)e 

7.5x101 ' 

(5.2x1 Q-1 p-remlg)g 

1.2x10-10 

(4.4x1 Qll p-rem/g)y 

4.Oxl Qll

7.Oxl 0`2 

5.5x1 Q-15 

2.6x1 Q-12 

6.2x1 0-13 

9.3x1 0-13 

4.2x1 Q-1 

1 .7x10'-3 

2.1x10-'3 

1 .2x1Q012 

1 .3x1 0-14. 

4.5x1 0-12 

5.lx10>'1 

(1. 1x 10-' p-rem/g)e 

6.7x1 10-' 

(4.6x1 Q10 p-rem/g)g 

6.4x10-10 

(2.4x1 Q100 p-rem/g)y 

2.4xl10 9

1 .5x10-" 

1 .5X10-13 

2.7x1 0-12 

1 .3x1 0-12 

2.8x10-" 

3.4x1 0-13 

5.0X10-12 . 

1 .1X10 13 

6.5x1 0-13 

3.4x1 0-13 

9.6X1 0-l2 

1 .9X10-10 

2.5X10-10 

(5.4x1 Qll p-rem/g)o 

2.8x`10-1l 

(1 .9x10>'1 p-rem/g)g 

2.7x1 011 

(1 .0x10 1' p-rem/g)' 

3.6x1 Q10

See following page for footnotes.

A.2-55
Footnotes to Table A.2.13

" Models and parameter values are presented in Appendix A.2.4.2. Except as noted for thorium 

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from 

disposal of a unit activity of 1 ,ACi (37 kBq) of radionuclides per year in all incinerators. DSRs for 

all exposure pathways should be reduced by a factor of 0.8 if a unit activity of 1 AzCi (37 kBq) of 

radionuclides per year is assumed to be disposed in landfills plus incinerators (see 

Appendix A.2.3.1.5 and A.2.4.2.5). 

b 1 person-rem/,uCi = 0.27 person-Sv/Bq. 

' Depending on the physical form of the particular items of concern, DSRs for exposure 

pathway could be reduced whenever radioactive materials should be significantly less 

dispersible than loose materials in trash (see Appendix A.2.3.1.5). 

d Calculated DSRs take into account the contributions from the decay products of 232

Th, which 

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and 

activity abundances of 232Th and 228Th in natural thorium are given in Table 3.1.1. 

e Value in units of dose per unit mass is based on the assumed activity abundances of 232Th 

and "

8Th and the specific activity of 232Th given in Table 3.1.3 (1 /MCi = 4.59 g). 

'Calculated DSRs take into account the contributions from the short-lived decay products of 21 U and 235

U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); 

assumed mass and activity abundances of 2-U, 235U, and 21U in natural uranium are given in 

Table 3.1.1. 

9 Value in units of dose per unit mass is based on the assumed activity abundances of 2

-8U, 21U, and 2

`U and the specific activities of uranium isotopes given in Table 3.1.2 

(1 yCi = 1.46 g). "h Calculated DSRs take into account the contributions from the short-lived decay products of 238U and 235

U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); 

assumed mass and activity abundances of 238U, 23

1U, and 2 4U in depleted uranium are given in 

Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 238U, 235U, and 2

34U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 uCi = 2.7 g).

A.2-56
Table A.2.14 Individual and Collective Dose-to-Source Ratios for Exposure to Off-Site 

Residents at Municipal Incinerators Due to Airborne Releases During 

Incinerator Operations a

Radionuclide 

3H 

14C 

36

CI 

46

Sc 

55

Fe 

60Co 

63Ni 

85

Kr 

9

`Sr + 90y 

99

Tc 

'

06Ru + 1°6Rh 

1°9Cd + lo9mAg 

1291 

1•Ba 

137Cs + l

37mBa 

147 Pm 

1

52 Eu 

204TI 

210mrBi + 21

6TI 

2 10po 

Th (natural)d 

U (natural)' 

U (depleted)h 

241

Am

Individual Effective Dose 

Equivalent 

(rem/,Ci)b 

1.0xl0-16 

8.7x1 0-19 

2.5x-10-15 

7.3x1 0-17 

1 .1x10-18 

1.6x10-15 

9.3x1 0-9 

5.6x10 19 

3.9x1 0-17 

5.Oxl 10-7 

7.5x1 0-17 

2.5x1 0-17 

3.1x10-1

5 

5.Oxl10-6 

1.1x10-1

6 

2.9x10- 18 

1.6x10-15 

9.4x1 0-18 

6.8x10-17c 

1.6x10-15 

1.7x10-

14 

(3.8x10-15 reim/g)e 

6.4x10-15 

(4.4x1 0-15 rem/g)9 

6.1x10-' 5 

(2.3x10-15 rem/g)' 

3.1x10-

14

See following page for footnotes.

A.2-57

Collective Effective Dose 

Equivalent 

(person-renh/p.Ci)b 

5.6x1 0-10 

1.lx10-11 

1.4xl 0-8 

2.9xl 0-10 

5.7x1 0-12 

5.9x10-9 

6.8x1 0-12 

4.6x1 0-12 

3.3x1 V0

4.7x 10-'0 

4.2x1 0-1 

2.0x10-lo10 

1.3x10-8 

1.8x10-9 

8.0x1 0-10 

1.9x10-11 

5.9x1 0-9 

4.5x10-11 

4.4x10-'oQ 

1.3x-10-8 

1.1x10-7 

(2.4x1 0-8 p-rem/g)e 

3.9xl 0-8 

(2.7xl 0-3 p-rem/g)g 

3.8xl 0-8 

(1.4xl 0-3 p-rem/g)i 

2.Oxl 0-7
Footnotes to Table A.2.14

a Models and parameter values are presented in Appendix A.2.4.3. Except as noted for thorium 

and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from 

disposal of a unit activity of 1 /Ci (37 kBq) of radionuclides per year in all incinerators. DSRs 

should be reduced by a factor of 0.2 if a unit activity of 1 /Ci (37 kBq) of radionuclides per year is assumed to be disposed in landfills plus incinerators (see Appendix A.2.3.1 .5 and A.2.4.3). b 1 person-rem//1 Ci = 0.27 person-Sv/Bq. 

c Value estimated based on the calculated DSR for natural thorium, which results primarily from 

inhalation exposure, and the assumption that the dose from 2

"OmBi also results primarily from 

inhalation exposure. 

d Calculated DSRs take into account the contributions from the decay products of 232 "Th, which 

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and 

activity abundances of 232Th and 28

Th in natural thorium are given in Table 3.1.1. 

e Value in units of dose per unit mass is based on the assumed activity abundances of 232Th 

and 228Th and the specific activity of 232Th given in Table 3.1.3 (1 YCi = 4.59 g). 'Calculated DSRs take into account the contributions from the short-lived decay products of 2

3U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); 

assumed mass and activity abundances of 23U, 235U, and 234U in natural uranium are given in 

Table 3.1.1. 

g Value in units of dose per unit mass is based on the assumed activity abundances of 238U, 2

`U, and 214U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 yCi = 1.46 g). 

" Calculated DSRs take into account the contributions from the short-lived decay products of 238U and 235U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); 

assumed mass and activity abundances of 2

31U, 235

U, and 2 4U in depleted uranium are given in 

Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 238U, 235U, and 214U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 /Ci = 2.7 g).

A.2-58
Table A.2.15 Individual Dose-to-Source Ratios for Exposure to Slag Workers 

at Metal Smelter or User of Automobile Containing Recycled Metal a 

Individual Effective

Radionuclide 

3

H 

14C 

38CI 

46Scf 

55Fe 

60

Co 

63Ni 

85Kr f 

'Sr + 9oY 

99

Tc 

106

Ru + 106

Rh 

109Cd + 1o9mAgf 

1291 

133 Baf 

'

37Cs + 137mtBa 

147pmn 

152Eu 

204Ti1

Dose Equivalent 

(rem//,tCi)b 

1.8x10-

14 

5.3x1 0-'3 

8.1x10-13 e 

1.6x10-13 e 

2.3x1 0-8 

1.9xl 0-13 

4.1x10-" 

3.9x1 0-13e 

2.3x1 0-9 

7.6x10-" 

5.4x10-9 

1.3x10-12 

1.lx10-8

Exposed Individualc 

Slag worker 

Slag worker 

Slag worker 

Slag worker 

Automobile user 

Slag worker 

Slag worker 

Slag worker 

Automobile user 

Slag worker 

Automobile user 

Slag worker 

Automobile user

Exposure Pathwayd 

Ingestion 

Ingestion 

Ingestion 

Ingestion 

External 

Ingestion 

Ingestion 

Ingestion 

External 

Ingestion 

External 

Inhalation 

External

21°rmBi + 206Tir 

210 Pof 

Th (natural)9 

U (natural)' 

U (depleted)'

2.2x1 0-8 

(4.8xl0-9 rem/g)h 

3.7x1 0-9 

(2.5x10-9 rem/g)j 

3.6x10-9 

(1.3x10-9 rem/g) 8 

1.7x10-8

Slag worker 

Slag worker 

Slag worker 

Slaa worker

See following page for footnotes.

A.2-59

Inhalation 

Inhalation 

Inhalation 

Inhalation
Footnotes to Table A.2.15

a Results are obtained from the previous analysis by Hill et al. (1995) as described in 

Appendix A.2.5. Except as noted for thorium and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents (EDEs) from disposal of unit activity of 1 ptCi (37 kBq) of 

radionuclides per year by metal recycling. 

b 1 rem/1tCi = 0.27 Sv/Bq. 

c Individual, either slag worker or automobile user, receiving the highest dose. 

SLimiting exposure pathway for determining dose to the slag worker or automobile user. e Hill et al. (1995) erroneously reported that external exposure to the automobile user was the 

limiting exposure pathway. Limiting exposure pathway was determined by analogy with the 

results for 9

°Sr, and the annual EDE was obtained by multiplying the result for 90

Sr by the 

ingestion dose coefficient for the radionuclide relative to the value for 9

°Sr. 

'Radionuclide was not included in analysis by Hill et al. (1995). 

g Calculated DSR takes into account the contributions from the decay products of 232Th, which 

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and the 

activity abundances of 23"Th and 228Th in natural thorium are given in Table 3.1.1. 

h Value in units of dose per unit mass is based on the assumed activity abundances of 232

Th 

and 28

Th and the specific activity of 232Th given in Table 3.1.3 (1 UCi = 4.59 g). 'Calculated 2 DSR takes into account the contributions from the short-lived decay products of 3

SU, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed 

mass and activity abundances of 238

U, 23"U, and 234U in natural uranium are given in Table 3.1.1. 

i Value in units of dose per unit mass is based on the assumed activity abundances of 23U, 235

U, and 2

34U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 ItCi = 1.46 g). 

k Calculated DSR takes into account the contributions from the short-lived decay products of 238U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed 

mass and activity abundances of 2 1U, 235U, and 234U in depleted uranium are given in 

Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 23

"U, 235U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 /Ci = 2.7 g).

A.2-60
Table A.2.16 Individual Dose-to-Source Ratios for Exposure to Off-Site Residents 

at Metal Smelter Due to Airborne Releases During Operations a 

Individual Effective Individual Effective 

Dose Equivalent Dose Equivalent 

Radionuclide (rem/IsCi)b Radionuclide (rem/gpCi)b 

3H 1.9x10- 16 1291 6.7x10-14 

14C 5.7x 01 8 1

33

Bac 

36c•c 137Cs + 137

m Ba 6.6x10-14 

46SCC 147Pm 3.3xl 0-16 

55Fe 1.7x10- 17 1

52Eu 7.7x10-14 

60

Co 7.4x10- 4 204

T1c 

63Ni 2.2xl 0-17 2

l1mBi + 2 6

Tlc 

85

Krc 210p o c 

90Sr +'Oy 2.9x1 0-1 Th (natural)d 2.9x1 0-12 

(6.3xl 0-13 reim/g)e 

99Tc 1.9x10- 16 U (natural)' 3.1x10-12 

(2.1 xl0-12 rem/g)g 

106

Ru + 1

06Rh 4.1x1 0-15 U (depleted)h 4.7xl 0-12 

(1.7xl 0-12 rem/g)i 

109Cd + l° 9 rnAgc 241Am 4.6x10-12 

a Results are obtained from the previous analysis by Hill et al. (1995) as described in 

Appendix A.2.5. Except as noted for thorium and uranium, dose-to-source ratios (DSRs) give annual effective dose equivalents from disposal of unit activity of 1 /Ci (37 kBq) of 

radionuclides per year by metal recycling. b 1 rem/ypCi = 0.27 Sv/Bq. 

c Radionuclide was not included in analysis by Hill et al. (1995). 

d Calculated DSR takes into account the contributions from the decay products of 232Th, which 

are assumed to be present and in activity equilibrium (see Table 3.1.3); assumed mass and 

activity abundances of 232Th and 2

8

Th in natural thorium are given in Table 3.1.1. e Value in units of dose per unit mass is based on the assumed activity abundances of 232Th 

and 228Th and the specific activity of 23 Th given in Table 3.1.3 (1 /Ci = 4.59 g). f Calculated DSR takes into account the contributions from the short-lived decay products of 23U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed 

mass and activity abundances of 238U, 21U, and 2 4U in natural uranium are given in Table 3.1.1. 

9 Value in units of dose per unit mass is based on the assumed activity abundances of 238U, "

235

U, and 2

`U and the specific activities of the uranium isotopes given in Table 3.1.2 

(11 uCi = 1.46 g). 

h Calculated DSR takes into account the contributions from the short-lived decay products of "

2 8U, which are assumed to be present and in activity equilibrium (see Table 3.1.2); assumed 

mass and activity abundances of 238U, 2

1U, and 21U in depleted uranium are given in 

Table 3.1.1. 

'Value in units of dose per unit mass is based on the assumed activity abundances of 238

U, 235

U, and 234U and the specific activities of the uranium isotopes given in Table 3.1.2 

(1 pCi = 2.7 g).

A.2-61
A.3 GENERIC DISTRIBUTION METHODOLOGY

A.3.1 Introduction 

Items that contain radioactive materials and are exempt from licensing requirements are very 

diverse in size, shape, and intended use environment. Therefore, distribution of exempt items 

can be accomplished by a variety of means. One or more items can be distributed from 

manufacturers to users by direct, nonstop commercial truck; by commercial package or mail 

delivery, which may involve truck and air transport and intermediate freight-handling terminals; 

and by wholesale and retail firms, which may involve all of the above plus warehouses, 

distribution centers, and retail stores. 

Methods and practices commonly used to distribute commercial and consumer goods have 

been characterized in a previous study (Etnier and O'Donnell, 1979). That study and personal 

observations of distribution equipment and facilities have been used to define a set of scenarios 

(i.e., typical distances and materials between people and packages and durations of exposure 

to the packages) that characterize routine distribution practices. (Potential accidents during distribution are addressed in Appendix A.1.) The defined scenarios, which are a small but 

representative subset of the many possible distribution scenarios, include: 

Commercial truck transport, which includes (1) nonstop (express) delivery via small, 

large, and tractor-trailer trucks; (2) local delivery via small and large trucks; and 

(3) regional or long-distance transport via small, large, and tractor-trailer trucks (see 

Appendix A.3.4). 

Warehousing, which includes handling in large warehouses (e.g., truck terminals) and 

medium-sized warehouses (e.g., distribution centers) (see Appendix A.3.5). 

Retailing, which involves handling, storage, and display in small and large retail 

establishments (see Appendix A.3.6). 

Air transport, which includes handling at receiving and shipping freight terminals and 

exposures to flight crew and passengers on a regularly scheduled flight (see 

Appendix A.3.7). 

The defined scenarios can be combined to build a model that is representative of most common 

distribution schemes (see Appendix A.3.3). Not all of the steps given in the scenarios may be 

needed for delivery of a specific item. In such cases, the modeler may choose to use only the 

appropriate parts of the scenarios. Also, item-specific analyses may be required for items that 

have the potential to cause high radiation doses or that are distributed by methods not 

characterized by the above. 

The defined scenarios and the methods described in Appendix A.3.2 are used to calculate 

individual and collective dose factors (DFs). A DF is the effective dose equivalent (EDE) 

(sievert (Sv) (rem)) associated with the distribution of an exempt item containing 1 microcurie 

(/,)Ci) (37 kilobecquerel (kBq)) of byproduct material or 1 mg of source material.

A.3-1
A.3.2 Derivation of the Dose Factors

Except for EDEs due to tritium (3

H), all EDEs associated with distribution are due to external 

exposures to radiation emitted from packages containing exempt items. These EDEs are 

calculated using a personal computer (PC) version of the mainframe CONDOS II methodology 

(Computer Codes, O'Donnell et al., 1981) and the exposure conditions given in Appendix A.3.4 

through A.3.7. The PC version performs the same calculations as the mainframe version, with 

the addition of the capability to calculate EDEs from internal and external exposures. Internal 

EDE calculations use the dose conversion factors discussed in Sections 2.1 and 3.1. External 

EDE calculations use the rotational geometry factors given in International Commission on 

Radiological Protection (ICRP) Publication 51 to convert dose in air (rad), which is the primary 

result of the CONDOS II calculations, to EDE (Sv (rem)). 

One source configuration is used in all external dose calculations, a 30-cm-long by 7.6-cm-radius concrete-like cylinder modeled as a self-absorbing cylinder with the dose point on 

the cylinder's axis, with or without external shielding (Computer Codes, O'Donnell et al., 1981). 

The source contains either 1 /Ci (37 kBq) of a byproduct material or 1 mg of a source material 

that is distributed uniformly throughout the cylinder. Because concrete has a density of 2.3 

g/cm3 and the volume of the cylinder is 5600 cm 3, the source mass is about 13 kg. Therefore, 

the source strength is 7.8x10-5 JCi/g (2.9 Bq/g) of byproduct material or 7.8x10-5 mg/g of 

source material. The atomic number of concrete, which is used in the calculation of dose due 

to bremsstrahlung, is 9.39. 

EDEs due to 3H exposure are calculated using Equation (4) of Appendix A.1.5 and the average 

airborne concentrations of 3H, which are given by 

C + = SV (e EtlV VC° ( e -EtIV 

E E 2 t Et 

where C = average airborne concentration of 3H (pCi/m3) during time t, 

t = time (h) over which C is averaged, 

Co = initial airborne concentration of 3H (pCi/m 3), 

S = source leak rate (pCi/h), 

E = volume ventilation rate (m3/h), and 

V volume (m3). 

Except for calculations involving exposures that begin after the leaking source has been 

removed (e.g., for a person along a truck route after the truck has passed), C. is set equal to 

zero (0). For persons along truck routes, separate calculations were performed for the period 

of vehicle passage and for the 8-hour period immediately after vehicle passage, for which CO 

was set equal to the final value of C during the period of passage. The source was assumed to 

contain 1 /Ci (37 kBq) of 3H and to have a fractional leak rate of 1 Pm/h; therefore, the value of 

S used in the calculations is 1x10`6 Ci/h (3.7x10 2 Bq/h). 

Tables A.3.1 and A.3.2 are lists of highly exposed and average individual DFs, by nuclide and 

delivery mode, for commercial truck transport of byproduct material. Table A.3.3 is a similar list

A.3-2
of collective DFs for truck transport of byproduct material (based on average driver conditions). 

Tables A.3.4, A.3.5, and A.3.6 are, respectively, lists of highly exposed and average individual 

DFs and collective DFs for truck transport of source material. Tables A.3.7 and A.3.8 are lists 

of individual and collective DFs, respectively, for air transport, warehousing operations, and 

retail sale of items containing byproduct or source materials. 

A.3.3 Application of the Dose Factors 

As illustrated below, application of the DFs requires the following six steps: 

Step 1. Identification and listing of the steps (scenarios) involved in the chosen 

distribution mode. 

Step 2. Identification and quantification of the radionuclides to be distributed. 

Step 3. Selection of highly exposed individual DFs, from Table A.3.1 or A.3.4 and 

Table A.3.7, and collective DFs from Table A.3.3 or A.3.6 and Table A.3.8. 

Step 4. Multiplication of the DFs by the quantity of radionuclide1 9 in the shipment to get 

individual and collective EDEs for each step. 

Step 5. Selection of highly exposed individual EDE for all steps. 

Step 6. Addition of the collective EDEs for all steps to get the total collective EDE for the 

shipment. 

To illustrate use of the DFs, consider delivery of 10 items, each containing 1 /zCi (37 kBq) of 

60

Co, via parcel delivery. There are several variations of parcel-delivery service. Some involve 

delivery only by truck and some involve delivery by truck and airplane. Considered here is the 

second delivery scheme, which requires 10 steps to move a shipment from a manufacturer to a 

retailer. 

Step 1. Transport from manufacturer to parcel-delivery center 1. 

Step 2. Process at parcel-delivery center 1. 

Step 3. Transport from parcel-delivery center 1 to air freight terminal 1. 

Step 4. Process at air freight terminal 1 and load airplane. 

Step 5. Transport from airport 1 to airport 2. 

Step 6. Unload airplane and process at air freight terminal 2. 

Step 7. Transport from air freight terminal 2 to parcel-delivery center 2. 

19 An additional step is required for 3H. If the assumed leak rate differs from 1 Pm/h, the DFs 

for 3H should be multiplied by the assumed leak rate, which must be expressed in units of Pm/h.

A.3-3
Process at parcel-delivery center 2.

Step 9. Transport from parcel-delivery center 2 to retailer. 

Step 10. Sell at retail store. 

As illustrated in Table A.3.9, Tables A.3.1 through A.3.8 are used to estimate a highly exposed individual EDE and a collective EDE for each of the above steps. Each step is represented by a 

scenario from the tables. Corresponding highly exposed individual DFs for 6

°Co are taken from 

Table A.3.1 or A.3.7 and multiplied by 10 (10 /Ci (370 kBq) are being shipped) to give a highly exposed individual EDE for each step. Step 10, Retailing - small store, gives the highest individual EDE (2x10-6 Sv (2xlO4 rem)) for this shipment. Similarly, collective DFs are taken 

from Table A.3.3 or A.3.8 and multiplied by 10 to give the collective EDE associated with each 

step. Scenario-specific EDEs then are summed to give the collective EDE associated with the 

shipment, which is 4x10 4 person-Sv (4x1 02 person-rem) in this example. 

A.3.4 Commercial Truck 

Essentially all distribution modes involve commercial truck transport. Three modes of truck 

transport-express, local, and regional delivery-can be used to create any desired 

truck-transport model. 

1. An express delivery is a short-distance, nonstop delivery between, for example, a 

manufacturer and a delivery service terminal or a delivery service terminal and a 

customer. A typical express delivery covers 48 km in 1 hour of driving-19 km in 

0.6 hour through high-population areas (3900 persons/km2

) and 29 km in 0.4 hour 

through medium-population areas (730 persons/km2). 

2. A local delivery covers the same distance as an express delivery but involves multiple 

stops at several delivery points. A typical express delivery covers 32 km and requires 4 hours, 1 hour driving and 3 hours making deliveries. Driving covers 13 km in 0.5 hour 

through high-population areas and 19 km in 0.4 hour through medium-population areas. 

The delivery truck is parked in high-population areas for the 3 hours required to make 

deliveries. 

3. Regional delivery is used to transport goods over long distances (e.g., between truck 

terminals). Over-the-road transport is a combination of regional deliveries that do not 

involve unloading cargo at intermediate terminals. A typical regional delivery, which may have one or multiple stops, covers 400 km and requires 5 hours of driving-20 km in 

0.7 hour through high-population areas, 20 km in 0.3 hour through medium-population 

areas, and 360 km in 4 hours through low-population areas (6 persons/km2

). 

A variety of truck sizes and shapes is available for use in commercial trucking. Since it is not 

possible to model every size and shape, three representative truck sizes (small delivery, large 

delivery, and tractor-trailer) are used in the assessment: 

A small delivery truck is taken to be a van-like vehicle with no structural barrier between 

driver and cargo. Typical dimensions are 305 cm long, 180 cm wide, 130 cm high, an

A.3-4

Step 8.
enclosed volume of 7.1x10 6 cm3, and an air-ventilation rate of 5 volume changes per 

hour. The driver sits 30 cm from the front and 180 cm from the center of the cargo. 

The truck body provides no shielding for the driver. 

A large delivery truck is taken to be a single-unit vehicle with no structural barrier 

between driver and cargo.2

" Typical dimensions are 500 cm long, 220 cm wide, 230 cm 

high, an enclosed volume of 2.5x10 7 cm3

, and an air-ventilation rate of 3 volume 

changes per hour. The driver sits 61 cm from the front and 340 cm from the center of 

the cargo. The truck body provides no shielding for the driver. 

A tractor-trailer rig typically is 1370 cm long, 235 cm wide, and 270 cm high, and has a 

cargo volume of 8.7x1 07 cm3 with an air-ventilation rate of 1 volume change per hour. 

The driver sits 140 cm from the front and 825 cm from the center of the cargo. The 

truck cab and trailer provide a 0.4-cm-thick aluminum shield for the driver. There is no 

air ventilation between the cab and trailer. 

Tables A.3.10 through A.3.12 provide the steps, types and numbers of persons exposed, and 

exposure conditions involved in shipment of goods via express, local, and regional truck 

delivery. 

A.3.5 Warehousing 

Except for direct delivery from a manufacturer to a customer or personal pickup by the 

customer, all methods of distribution involve one or more warehouse-like facilities. These 

facilities may include commercial truck terminals, large warehouses (chain-store or direct 

merchandisers), and smaller distribution centers or stockrooms in retail establishments. In 

such facilities, one or more two-person crews load and unload trucks, a forklift operator moves 

pallets of material, and one or more sorters direct incoming packages to the appropriate outgoing truck bay. Two warehouse-like facilities are considered in this appendix: 

A large warehouse is taken to be about 4600 cm long, 3050 cm wide, and 610 cm high, and to have an enclosed volume of 8.5x10 9 cm3 with an air-ventilation rate of 1 volume 

change per hour. Each of five truck bays has a volume of 3.0x108 cm3 with an air

ventilation rate of 4 volume changes per hour. 

A medium-sized warehouse is taken to be about 3050 cm long, 3050 cm wide, and 

370 cm high, and to have an enclosed volume of 3.4x109 cm3 with an air-ventilation rate 

of 1 volume change per hour. Each of three truck bays has a volume of 3.0x10 8 cm3 

with an air-ventilation rate of 4 volume changes per hour. 

The steps, types and numbers of workers, and exposure conditions involved in handling goods in large and medium-sized warehouse-like facilities are given in Tables A.3.13 and A.3.14, 

respectively. 

20 Many trucks fitting this definition of large have, in fact, separate driver and cargo 

compartments. For the sake of being reasonably conservative, the case of joined 

compartments is used.

A.3-5
A.3.6 Retailing

Exempt items that are designed for use by members of the public are frequently distributed 

through retail stores. In such stores, the items are put on display and are near and handled by 

sales clerks and customers. Two retail store sizes, large and small, are modeled in this 

appendix: 

A large retail store is taken to be about 3050 cm long, 1520 cm wide, and 610 cm high, 

and to have an enclosed volume of 2.8x10 9 cm3 with an air-ventilation rate of 4 volume 

changes per hour. 

A small retail store is taken to be about 1520 cm long, 910 cm wide, and 460 cm high, 

and to have an enclosed volume of 6.4x10 8 cm3 with an air-ventilation rate of 4 volume 

change per hour. 

The steps, types and numbers of persons, and exposure conditions involved in handling goods 

in large and small retail stores are given in Tables A.3.15 and A.3.16, respectively. 

A.3.7 Air Transport 

Distribution of many smaller packages involves air transport. Typically, air cargo is delivered to 

an air-freight terminal, where it is unloaded, sorted, and put into containers. (These steps may 

be performed by airport or shipper's personnel.) Containers are transported to and loaded on 

an airplane, which transports the cargo to another airport, where the above steps are reversed. 

During the flight, exposed persons include the flight crew, attendants, and passengers: 

An air-freight terminal is taken to be about 3050 cm long, 3050 cm wide, and 610 cm 

high, and to have an enclosed volume of 5.7x109 cm3 with an air-ventilation rate of 

1 volume change per hour. Each of three truck bays has a volume of 3.0x10 8 cm3 and 

an air-ventilation rate of 4 volume changes per hour. 

The cabin area of an airplane is taken to be about 1800 cm long, 460 cm wide, and 

305 cm high, and to have an enclosed volume of 2.6x10 8 cm 3 with an air-ventilation rate 

of 3 volume changes per hour. Air freight normally is stowed in two baggage 

compartments, a forward and an anterior hold, located below the cabin area. 

The steps, types and numbers of persons, and exposure conditions involved in air transport of 

goods are given in Tables A.3.17 and A.3.18.

A.3-6
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Table A.3.2 Average (Package In Center of Cargo Area) Individual DFs 

for Commercial Truck Transport of Byproduct Material a

--Express Delivery Local Delivery Regional Delivery 

Radionuclide Small Truck Large Truck Semi-Truck Small Truck Large Truck Small Truck Large Truck Semi-Truck 

11Mg2.8xl 0-7 2.1xl X1- 1.4x 0-7 6.3x 0-7 5.0x 0-7 3.7x 0-7 7.4x108

' 4.8x-10-9 

2Am6.OxlQ 10 4.6x 10`0 3.lxlO 10 1.3xl10 9 1.lXlO 9 7.4x10 10 1.5x 10-'0 8.2xl 0-2 

13

'Ba 3.4x-10-8 2.5x10-8 1.7x-10-8 7.5x10 8 6.OxlO-8 4.5x10-1 8.9X10-9 5.7x10-10 

140 1.2x10>-'2 8.5x 0-13 6.Ol0-103 2.6xl 0-12 2.1 Xlo 12 1.5x 0-12 3.2x10-'3 2.2x10'-4 

1090d 1.8x10 10 1.4x10 10 9.7x10- 1 4.1x10-10 3.5xl10'0 2.4x10-10 5.3x10 " 3.9xl 0-2 

36C! 1.9x10-10 1.4x1010

o 9.6xl011 4.2x10 10 3.4x1010 2.5xl1010 5.lxlO11 3.3x10>-12 

6000O 2.7x1 0-7 2.OxlO 0- 1,4x10-7 6.0x10-7 4.9x1 0-7 3.6x 0-7 7.3x 0-8 5.OxlO 9 

137cs 6.1 X10-8 4.4x10-8 3.0x 0-8 1.3x1 0-7 1.OXlO 7 7.9x 0-8 1.6x10-8 9.7x10 10 

3 H 2.1lX1 O 12 1.2x1 0-12 4.3x 0-13 6.4x1 0-12 3.9x 0-12 1.2x10-" 5.4x10'-2 O.0X10-0 

85 Kr 4.4x10-10 3.2x10'10 2.2xl1010 9.6x10-10 7.6xl10'0 5.7x10-10 1.lXlO 10 7.2x10'-2 

63 Ni 3.5x 0-14 2.7x1 0-14 1.9X10-14 7.5x1O0-14 6.0x10-14 4.3'xl-14 8.5xlO0-15 5.5x 0-16 

147 Pm 2.5x1 0-12 1.9X10-12 1.3x1 0-12 5.5x1 0-12 4.5x1 0-12 3.2x10'-2 6.7x10-'3 4.5x1 0-14 

2PO7.8x 0-13 5.7x 0-13 3.8x10 -'3 1.7xl 0-12 1.3x 0-12 1.0X10-12 2.OxlO 13 1.3x10 -14 

46sc 2.1 X10-7 1.6x 0-7 1.0X10-7 4.6x 0-7 3.7x 0-7 2.7x 0-7 5.5x10-8 3.6x10 9 

90Sr 1.7x1010

o 1.3xl1010 8.6xl011 3.8x10-'0 3.0x10-'0 2.2x10-'0 4.5xl10" 3.0x 0-12 

99Tc 1.2xl1010 8.7x1011 5.9x1011 2.6x10 10 2.1 x10-'0 1.5x10-10 3.lxl011 2.1 X10' 2 

aUnits are in rem//pLCi shipped. 1 rem/ktCi =0.27 Sv/Bq.

A.3-8
Table A.3.3 Collective DFs for Commercial Truck Transport 

of Byproduct Material I

Express Delivery Local Delivery Regional Delivery 

Radionuclide Small Truck Large Truck Semi-Truck Small Truck Large Truck Small Truck Large Truck Semi-Truck 

ilomAg 3.0x10-7 2.2x 0-7 1.4x1 0-7 7.3x 0-7 5.4x 0-7 3.9x 0-7 8.3xl108 6.2xl10 9 

24An6.3x 101'

0 4.7x 10 3.x1 1WO 1.5X10-9 1.2x109' 7.7x10-10 1.6xl10'0 1.1X10 11 

133

13a 3.7x 0-8 2.6x10 8 1.7x108

" 8.8x 0-8 6.5x10 8 4.7x 0-8 9.8xl10 9 7.2xl 0l0 

1C1.3x1 0-12 9.0x10-13 6.0x10<-'3 3.OxlQ -12 2.3x10'-2 1.6x10-'2 3.5xl 0-13 2.9x1 0-14 

19d2.0x10- 10 1.5x10-10 9.9xI10' 1 4.9x10-10 3.910-10 2.6x10-10 6.0xl0'1 5.6x1 0-12 

36a1 2.lxlO 10 1.5x10-'0 9.7x1011 5.OxlO-10 3.7x 101 '

0 2.6x10 10 5.6x10 " 4.3x10>' 2 

60CO 2.9x1 0-7 2.1 X10-7 1.4x 0-7 7.lXlO 7 5.4x 0-7 3.8x 0-7 8.2x108" 6.6x 0-9 

137Cs 6.5x10-8 4.6x 0-8 3.0x108

' 1.6x1i0- 1.1X10-7 8.3xl108 1.7x 0-8 1.2x 0-9 

3H2.1 X10-12 1.2x 0-12 4.3x100-13 6.4x10'-2 3.9x10>-'2 1.2x10<'1 5.4x1 0-12 00OX10-0 

85 Kr 4.7x10-10 3.3xl1010 2.2x10-10 1.1X10-9 8.3x10 10 6.Ox10O>'0 1.3x10 10 9.2x10-12 

63Ni 3.7x 0-14 2.8x1 0-14 1.9X10-14 8.5x10'-4 6.5x10'-4 4.5x 0-14 9.5X 10- 15 8.Ox 0-16 

1

47Pm 2.7x10>-12 1.9X10-12 1.3x 0-12 6.5x 0-12 4.9x 0-12 3.4x10'-2 7.3x 0-13 6.1 X10-14 

210PO 8,3x10-'3 5.9x10>-'3 3.9x1 0-13 2.OxlO -12 1.5X10-11 1.1 X10-12 2.2xl 0-3 1.6x1 0-14 

46sc 2.3x1 0-7 1.6x 0-7 1.lXl Q7 5.4x 0-7 4.OxlO 0- 2.9x1 0-7 6.lx108" 4.6xl10 9 

90Sr 1.8x1010

o 1.3x10-10 8.7x101' 4.5x1010o 3.3x1 10- 2.4x 10` 0 5.1X10-11 3.9x1 0-12 

99Tc 1.3x10-'0 9.1X10-11 6.OxlO " 3.OxlO 10 2.3x10-10 1.6x10- 10 3.5x10'1 2.7x10'-2

a Units are in person-rem//pCi shipped. I person-rem/mCi = 0.27 person-Sv/Bq.

A.3-9
Table A.3.4 Highly Exposed (Package Near Driver) Individual DFs for Commercial Truck Transport of Source Material" 

Express Delivery Local Delivery Regional Delivery 

Radionuclide Small Large Truck Semi-Truck Small Truck Large Truck Small Truck Large Truck Semi-Truck 

Truck 

naturaiTh (1 yr) 1.1x10-10 3.Ox1O-" 9.5x10- 12 2.2x10-l' 6.6x10- 11 4.9x10-<' 1.0x10-1° 9.6x1O0-12 

nat

uraITh (20 yr) 2.0x10-1 ° 5.3x10-" 1.7x10 -1 4.OxlO10- 1.2x10-10 8.8x10-'0 1.8x10-10 1.6x10-" 

depletedu 7.1x10-' 2 1.9x10-12 5.8x10- 13 1.4x10-11 4.1xi0-12 3.2x10-11 6.2x10-12 5.2x10-13 

naturalu 8.1x10- 12 2.2x10-12 6.6x10-1 3 1.6x10- 4.7x10- 12 3.7x10-" 7.1x 10-12 6.OxlO-13

a Units are in rem/mg shipped. 1 rem/mg = 0.01 Sv/mg.

A.3-10
Table A.3.5 Average (Package in Center of Cargo Area) Individual DFs 

for Commercial Truck Transport of Source Material a

Radionuclide 

naturafTh (1 yr) 

naturalTh (20 yr) 

depletedu 

naturalu

Small Truck 

1.5x10 -1 

2.6x10

9.4x10-13 

1.0x10-1

2

Express Delivery 

Large Truck S 

1 .lx10-1 

2.OxlO-" 

7.Oxl 0-13 

8.0x10-1

3

emi-Truck 

7.6x10-12 

1.3x10 -1 

4.8x101-3 

5.5x1 0-13

Local Delivery 

Small Truck Large Truck 

3.2x10- 2.7x10

5.8x10-1 4.8x10-" 

2.1x10-1 2 1.7x10-12 

2.4x10-12 1.9x10- 12

Regional Delivery 

Small Truck Large Truck Semi-Truck 

1.9x10-11 4.1x0O- 12 3.1x10-' 3 

3.4x10'11 7.2x10-12 5.3x10-13 

1.2x10- 12 2.5x10- 13 1.7x10-14 

1.4x10-12 2.9x10l1 3 2.Oxl 0-14

a Units are in rem/mg shipped. 1 rem/mg = 0.01 Sv/mg.

A.3-11
Table A.3.6 Collective DFs for Commercial Truck Transport of Source Material a

Radionuclide 

naturaiTh (1 yr) 

nat

uralTh (20 yr) 

depletedu 

naturalu

SmallTruck 

1.6x10

2.8x10- " 

1.0xl 0-12 

1.2x10-12

Express Delivery 

Large Truck Semi-Truck 

1.2x10-1 7.7x10- 12 

2.1x10-11 1.4x10-•1 

7.3x10'- 3 4.8x10-13 

8.4x10-' 3 5.5x10- 13

Local Delivery 

Small Truck Large Truck 

3.9x10- 3.0x10-" 

6.9x10-" 5.3x10-11 

2.5x10- 12 1.8x10-12 

2.8x10-1 2 2.1xlo-1 2

Small Truc 

2.OxlO1 - 1 

3.6x10-" 

1.3x10-12 

1.5x10-12

Regional Delivery 

¢k Large Truck S 

4.7x1 0-12 

8.1x10-12 

2.8x10-1 3 

3.2x10-13

' Units are in person-rem/mg shipped. 1 person-rem/mg = 0.01 person-Sv/mg.

A.3-12

emi-Truck 

4.5x10-13 

7.3x1 0-13 

2.2x1 0-14 

2.6x10-14
e BuillejoU pue'BuisnotlejeAh I:podsuoi.L jiV jol s:ja jenPIAIPUJ pasodx3 AILIBIH LcV e1qej.

c I-C'V

,,-OLX9.L 

It-OLXS'9 

6-0 IX9. ý 

6-0 VO- L 

6-OVL-8 

9-0 Ln. L 

9-0 LX9. L 

tt-OLXq-q 

0,-0 LXL. L 

,-Ovuz 

O-Ovvc 

6-0 VET 

tt-OVL*9 

9-0 V81, 

9-0 M'L 

9-OIX61 

9-0 Lxz;. L 

tL-Ovo.q 

9-oLxt,.z 

8-0 LXL*C 

9-0 l.XOZ 

OJOIS 1jews

,I-O[XL-9 

II-Ovo-9 

6-0 VV ý 

Ot-OvL-L 

6-OV6-9 

6-OLX3'6 

S-OLX[*[ 

11-0VTV 

0,-0 Lxc. L 

zi-MWL 

s-OVtI*z 

6-01XOT 

t t-0 LX2. L 

9-ovc-c 

5-0 Vs, L 

8-0 vo. i. 

6-OLX9'6 

LL-OLXO'9 

9-OIX91 

a-OLXZ'C 

9-0 Vs, L 

OJOIS GBAB-1

3t-Ovz-v ct-MV8 

3,-OLXL*c et-ovc-L 

ot-ol.XrL It-Ovrz 

tL-OLXO'9 1, -0 lx?,. L 

Ot-WS't, wOLXL-6 

Ot-OV9,9 Ot-OM, ý 

L-0 V3,8 L-0 I.Xf L 

zt-ovo-c c'-w rg 

zt-OVEZ ,-OLX6' 1. 

C,-Ovz-ý pt-OLX9.3 

6-0 LXL* L 0,-OLX17.6 

Ot-OLn-L t,-OLXO*C 

at -0 Lxo. L CI-OIXO*B 

L-0 LXC*Z 8-0 VL't, 

9-MLI L-OLXZ.z 

o,-OVUL 04 Vs, L 

o,-OLX6'9 OL-01.xt'I 

zt-OLxt,*t, ct-OVO-6 

L-0 LXC* L 8-wxLz 

6-0 VO,2; ot-ovvt, 

g-OvL-L L-0 LXZ*Z 

asnotimAA wn1poW 99=49jeM sOje-1 

Buisnoi4exAA

zt-OIX9*z 

3t-Ol.XC'Z 

11-0[xt,.g 

it-ovq-c 

Ot-Ov.8-3 

0ý-Olxz:-t, 

,-0 vz-q 

z,-O V6- L 

at-01.Wg 

II-O[XO'L 

6-OVVI 

tt-OM's 

C,-OVt"z 

L-0 V9. l' 

,-OLXL'9 

ol-OIXL't, 

01-OVE't, 

a,-O[xq-z 

8-01xv-8 

6-0 VZ. L 

L-0l.XO'L 

ouLIdjlgf

CI-0i.xv-S 

cj-Oi.XL-t, 

j,-O [xc. L 

z,-OLXS'L 

II-OlX8.9 

tj-O[Xq-q 

L-OLXO.L 

c'-wo-C 

Z,-O m. L 

Ot-OIX8.1 

0,-0 LX rz 

II-OVIVE 

zt-0 [xv. L 

8-ovo-c 

,-0 [xv. 1 

j,-Ol.x9-6 

,,-OLX9'6 

C,-OLXO.g 

0-0 LXL. I 

Ot-Ovve 

L-0 [Xt* ý 

IBUIWAO.L ILIBIGAZI

qnlejnieu 

qnp,,,,dp 

q(jAOE;) Uleinjeu 

q(JA qilLjnlvu 

OJ-66 

JS06 

OS90 

OdoLz 

WCALK 

!Ncg 

J>199 

IM 

He 

SOM 

0009 

loge 

P0601 

opt 

L2ecct 

WVLPZ 

6vwOI L 

eplIonuoilpeU

'15w/big Loo = 6w/wei L -6w/wei ale swnq L3'O =!Ori/WOJ L 'peddiqs iorf/wei u, ale si!un,,

isumelou liodsuej.L jIV
Table A.3.8 Collective DFs for Air Transport, Warehousing, and Retailing I

Radionuclide 

ilom"Ag 

241

Am 

133 Ba 14C 

109Cd 

36CI 

60C0 

3 H 

1291 

85 Kr 

63 Ni 

210

OP0 

46 SC 

90Sr 

99

Tc 

natura~l h (1 yr)b 

nat

uralrTh (20 yr)b 

depletedub 

naturalub

a Units are in person-rem//pCi shipped. 1 person-rem/p~i = 0.27 person-Sv/Bq. ' Units are rem/mg. 1 rem/mg = 0.01 Bq/mg.

A.3-14

Air Transport 

Freight Terminal Airplane 

7.1 X10-7 1.4x1 0

1.4x1 0-9 1.9x 0108 

8.4x10 8

I 1.7x1 0-6 

3.0x101 2 5.5x1011l 

4.8x101 0 8.8x10-9 

4.8xl101 0 9.5x 0-9 

6.9x1 0-7 1.4x 0-5 

1.5x 0-7 3.0x 0-6 

3.1 X1012 3.lx101' 

1.1x10-10 1.3x10-9 

1.1X10-9 2.2x 0-8 

8.5x 0-14 1.3x 0-'2 

6.3x10-12 1.2x10-'0 

1.9X10-12 3.9x101' 

5.2x 0-7 1.1x105

, 

4.3x10-10 8.5x10-9 

2.9x10 10 5.8x 0-9 

3.910-11 7.8x10-10 

6.8x1011 1.4x109' 

2.4x1 0-12 4.7xO101 

2.7x10 12 5.4x10 lI

Warehousing 

Large Warehouse Medium Warehouse 

1.6x1 0- 6.4x10-6 

3.1 x10-9 1.2x10 68 

2.OxlO 0- 7.6x10-7 

5.5x10'-2 3.3x10'1 

1.1X10-9 4.1 x10-9 

1,1X10-9 4.3x1 0

1.6x1 0-6 6.3x 0-6 

3.4x10-7 1 .3x106

' 

3.8x10-12 1.5X10-11 

2.3x1 Q10 8.5x1 10-' 

2.5x109' 9.7x 0-9 

1.8X10-13 7.0x10-13 

1.4x1011I 5.5x10-11 

4.4x 0-12 1.7x10- 11 

1.2x1 0-6 4.8x1 0-6 

9.8x1 010 3.8x1 0-9 

6.7x10 10 2.6x1 0

9.1X10-11 3.5x10100 

1.6x10 10 6.2x10-10 

5.5x1 0- 2 2.1x1011' 

6.3x10-12 2.4x10 11

Large Stor 

2.3x1 0-3 

4.1 X10-6 

2.6xl 0" 

1.1x10- 6 

2.Oxl 06 

1.6x 0-6 

2.4x1 0-3 

4.4x 0-4 

4.6x10 8 

2.3x1 0-7 

3.3x1 0-6 

2.8x1 QlO 

2.2x1 0-8 

5.8x1 0lI 

1 .7x 0-3 

1 .4x1 0-6 

9.8x1 0-7 

1 .6x1 0-7 

2.6x1 0-7 

8.Oxl 0-9 

9.4x1 0-9

Retailing 

e Small Store 

3.8x1 O-3 

6.9x1 0-6 

4.5x 10'

1.7x1 0

3.Oxl 0-6 

2.6x1 0-6 

3.9x1 0-' 

7.8x1 0-4 

2.Oxl 0-8 

3.9x1 0-7 

5.7x1 0-6 

4.4x1 010 

3.6x1 0-8 

I.0X10-8 

2.8x1 0-3 

2.4x1 0-6 

1.6x1 0-6 

2.4x1 0-7 

4.1 X10-7 

1 .3x1 0-8 

1.5X10-8
Table A.3.9 Example Estimation of Individual and Collective Effective Dose Equivalents (EDEs) 

Highly Exposed Individual Collective 

DF EDE DF EDE Step Scenario Used Source Table (rem/,pCi)a (rem)a Source Table (person-rem/gCi)O (person-rem)' 

1 Express delivery A.3.1 2.Oxl 0-6 2x10- A.3.3 2.9x10- 7 3x10-6 

Small truck 

2 Warehousing A.3.7 2.2x10-7 2x10-6 A.3.8 1.6x10- 6 2x10

Large warehouse 

3 Express delivery A.3.1 5.5X10-7 6x10-6 A.3.3 2.1x10-7 2x10-6 

Large truck 

4 Air transport A.3.7 1.4x10-7 1x10-6 A.3.8 6.9x10- 7 7x10-6 

Freight terminal 

5 Air transport A.3.7 6.7x10-7 7x10-6 A.3.8 1.4x10-5 1x10-4 

Airplane 

6 Air transport A.3.7 1.4x10- 7 lx10-6 A.3.8 6.9x10-7 7x10-6 

Freight terminal 

7 Express delivery A.3.1 5.5x10-7 6x10-6 A.3.3 2.1x10-7 2x10-6 

Large truck 

8 Warehousing A.3.7 2.2x10-7 2x10- 6 A.3.8 1.6x10- 6 lx10-5 

Large warehouse 

9 Local delivery A.3.1 4.1x10-6 4x10-5 A.3.3 7.1x10- 7 7x10-6 

Small truck 

10 Retailing A.3.7 1.9x10 5

- 2x10-4 A.3.8 3.9x10 3 4x10-2 

Small store 

a 1 rem/yCi = 0.27 Sv/Bq; 1 person-rem/gCi = 0.27 person-Sv/Bq. 1 rem = 0.01 Sv; 1 person-rem = 0.01 person-Sv. 

A.3-15
Table A.3.10 Steps, Types and Numbers of Persons Exposed, and Exposure Conditions 

Involved in Shipment of Goods Via Express Truck Delivery

Absorbers 

Exposure Concentration Exposure 

Exposure Time of 3H in Air Distance Material Thickness 

Event (h) (pCi/m 3

)a (cm) (cm) 

SMALL TRUCKS 

1 driver 

Driving 

- high 1.0 0.016 30 Wood 15 - average 1.0 0.016 180 Wood 15 

Handling 0.033 0.016 30 Wood 1.0 

In truck 0.3 0.016 90 Wood 15 

Near truck 0.66 0.00011 210 Wood 30 

Aluminum 0.32 

120,000 persons in hiqh-population zones 

Along route 0.050 7.7x 10-'0 18,300 Wood 30 

Aluminum 0.32 

8.0 3.1x10-11 

34,200 persons in medium-poDulation zones 

Along route 0.022 3.6x10-'0 18,300 Wood 30 

Aluminum 0.32 

8.0 1.5x10-11 

a 1 pCi/m 3= 0.037 Bq/m 3.

A.3-16
Table A.3.10 Steps, Types and Numbers of Persons Exposed, and Exposure Conditions 

Involved in Shipment of Goods Via Express Truck Delivery (continued) 

Absorbers 

Exposure Concentration Exposure 

Exposure Time of 3H in Air Distance Material Thickness 

Event (h) (pCi/m3 )a (cm) (cm) 

LARGE TRUCKS 

1 driver 

Driving 

- high 1.0 0.0091 60 Wood 30 - average 1.0 0.0091 340 Wood 30 

Handling 0.033 0.0091 30 Wood 1.0 

In truck 0.50 0.0091 90 Wood 30 

Near truck 1.5 0.00018 310 Wood 60 

Aluminum 0.32 

120,000 persons in high-population zones 

Along route 0.050 6.5x10-10 18,300 Wood 60 

Aluminum 0.32 

8.0 2.6x10"

34,200 persons in medium-population zones 

Along route 0.022 3.0xl0-10 18,300 Wood 60 

Aluminum 0.95 

8.0 1.2x10"

a 1 pCi/m 3 = 0.037 Bq/m 3.

A.3-17
Table A.3.1 0 Steps, Types and Numbers of Persons Exposed, and Exposure Conditions 

Involved in Shipment of Goods Via Express Truck Delivery (continued) 

Absorbers 

Exposure Concentration Exposure 

Exposure Time of 3H in Air Distance Material Thickness 

Event (h) (pCi/m3)a (cm) (cm) 

TRACTOR-TRAILER TRUCKS 

1 driver 

Driving Wood 60 

- high 1.0 0.0 140 Aluminum 0.95 

- average 1.0 0.0 830 Wood 60 

Aluminum 0.95 

Handling 0.033 0.00016 30 Wood 1.0 

In trailer 1.0 0.0037 120 Wood 60 

Near trailer 3.0 0.00025 460 Wood 120 

Aluminum 0.32 

120,000 persons in high-population zones 

Along route 0.050 3.5xl 0-'0 18,300 Wood 120 

Aluminum 0.32 

8.0 1.4x10- 11 

34,200 persons in medium-population zones 

Along route 0.022 7.7x10-10 18,300 Wood 120 

Aluminum 0.32 

8.0 6.7x10-12

a1 pCi/m3 = 0.037 Bq/m3 .

A.3-18
Table A.3.11 Steps, Types and Numbers of Persons Exposed, and Exposure 

Conditions Involved in Shipment of Goods Via Local Truck Delivery 

Absorbers 

Exposure Concentration Exposure 

Exposure Time of 3H in Air Distance Material Thickness 

Event (h) (pCi/m3)a (cm) (cm) 

SMALL TRUCKS 

1 driver 

Driving 

- high 2.0 0.018 30 Wood 15 - average 2.0 0.018 180 Wood 15 

Handling 0.33 0.0016 30 Wood 1.0 

In truck 1.0 0.016 90 Wood 15 

Near truck 3.0 0.000078 210 Wood 30 

Aluminum 0.32 

80,000 oersons in high-population zones 

Along route 0.47 3.8xl 0-9 18,300 Wood 30 

Aluminum 0.32 

7.5 1.2x10-10 

22,800 persons in medium-population zones 

Along route 0.022 3.6x10-'0 18,300 Wood 30 

Aluminum 0.32 

8.0 1.5x10-'1

a 1 pCi/m 3 = 0.037 Bq/m 3.

A.3-19
Table A.3.11 Steps, Types and Numbers of Persons Exposed, and Exposure 

Conditions Involved in Shipment of Goods Via Local Truck Delivery (continued) 

Absorbers 

Exposure Concentration Exposure 

Exposure Time of 3H in Air Distance Material Thickness 

Event (h) (pCi/m 3)a (cm) (cm) 

LARGE TRUCKS 

1 driver 

Driving 

- high 2.0 0.011 60 Wood 30 

- average 2.0 0.011 340 Wood 30 

Handling 0.033 0.00012 30 Wood 1.00 

In truck 2.0 0.011 90 Wood 30 

Near truck 3.0 0.000079 310 Wood 60 

Aluminum 0.32 

80,000 persons in high-population zones 

Along route 0.47 3.9xl 0-9 18,300 Wood 60 

Aluminum 0.32 

7.5 1.2x10 0

-1 

22,800 persons in medium-population zones 

Along route 0.022 3.7x10-'° 18,300 Wood 60 

Aluminum 0.32 

8.0 1.5x10"-

a 1 pCi/mr3 = 0.037 Bq/m 3.

A.3-20
Table A.3.12 Steps, Types and Numbers of Persons Exposed, and Exposure 

Conditions Involved in Shipment of Goods Via Regional Truck Delivery 

Absorbers 

Exposure Concentration Exposure 

Exposure Time of 3H in Air Distance Material Thickness 

Event (h) (pCi/m3 )a (cm) (cm) 

SMALL TRUCKS 

1 driver 

Driving 

- high 5.0 0.019 30 Wood 15 - average 5.0 0.019 180 Wood 15 

125,000 persons in high-population zones 

Along route 0.050 9.2xl0-10 18,300 Wood 30 

Aluminum 0.32 

8.0 3.7xl0-1 

23,750 persons in medium-population zones 

Along route 0.022 4.3x10-'0 18,300 Wood 30 

Aluminum 0.32 

8.0 1.7x10-'1 

4,500 persons in low-population zones 

Along route 0.015 3.0x10'-0 18,300 Wood 30 

Aluminum 0.32 

8.0 1.2x10-1 

a 1 pCi/m 3= 0.037 Bq /M 3.

A.3-21
Table A.3.12 Steps, Types and Numbers of Persons Exposed, and Exposure 

Conditions Involved in Shipment of Goods Via Regional Truck Delivery (continued) 

Absorbers 

Exposure Concentration Exposure 

Exposure Time of 3H in Air Distance Material Thickness 

Event (h) (pCi/m3)a (cm) (cm) 

LARGE TRUCKS 

1 driver 

Driving 

- high 5.0 0.012 60 Wood 30 - average 5.0 0.012 340 Wood 30 

125,000 persons in high-population zones 

Along route 0.050 8.9x10-'0 18,300 Wood 60 

Aluminum 0.32 

8.0 3.6x10-11 

23,750 persons in medium-population zones 

Along route 0.022 4.2xl0-° 18,300 Wood 60 

Aluminum 0.32 

8.0 1.7x10"l 

4,500 persons in low-population zones 

Along route 0.015 2.9x10-10 18,300 Wood 60 

Aluminum 0.32 

8.0 1.2xl0-"

a 1 pCi/m 3 = 0.037 Bq/m 3 .

A.3-22
Table A.3.12 Steps, Types and Numbers of Persons Exposed, and Exposure 

Conditions Involved in Shipment of Goods Via Regional Truck Delivery (continued) 

Absorbers 

Exposure Concentration Exposure 

Exposure Time of 3H in Air Distance Material Thickness 

Event (h) (pCi/m3)a (cm) (cm) 

TRACTOR-TRAILER TRUCKS 

1 driver 

Driving 

- high 5.0 0.0 140 Wood 60 

830 Aluminum 0.95 

- average 5.0 0.0 Wood 60 

Aluminum 0.95 

125,000 persons in hiqh-population zones 

Along route 0.050 7.7xl 0-0l 18,300 Wood 120 

Aluminum 0.32 

8.0 3.1x10-11 

23,7500 persons in medium-population zones 

Along route 0.022 3.6x10-'0 18,300 Wood 120 

Aluminum 0.32 

8.0 1.5x10-11 

4,500 person in low-population zones 

Along route 0.015 2.5x10-10 18,300 Wood 120 

Aluminum 0.32 

8.0 1.0xl0-11

" 1 pCi/m3 = 0.037 Bq/m3.

A.3-23
Table A.3.13 Exposure Conditions for a Large Warehouse

Absorbers 

Exposure Concentration Exposure 

Exposure Time of 3H in Air Distance Material Thickness 

Event (h) (pCi/m3)a (cm) (cm) 

4 PRIMARY LOADERS AND UNLOADERS 

Handle cargo 0.017 0.0088 30 Wood 1.0 

In trailer 1.0 0.0088 120 Wood 60 

In trailer bay 2.0 0.00073 460 Wood 120 

Aluminum 1.0 

16 OTHER LOADERS AND UNLOADERS (2 PER BAY) 

1 bay away 2.0 0.00018 910 Wood 230 

Aluminum 1.0 

2 bays away 2.0 0.0000071 1,400 Wood 340 

Aluminum 1.0 

3 bays away 2.0 0.000011 1,800 Wood 450 

Aluminum 1.0 

4 bays away 2.0 0.0000029 2,300 Wood 550 

Aluminum 1.0 

5 FORK LIFT OPERATORS 

Move cargo 0.050 0.00073 120 Wood 1.0 

Iron 0.16 

Other work 5.0 0.000094 1,000 Wood 250 

Aluminum 10 

Iron 1.0 

10 STOREROOM CLERKS 

Handle cargo 0.033 0.000094 30 Wood 1.0 

Near cargo 1.0 0.000094 120 Wood 30 

Other work 4.0 0.000094 610 Wood 15 

Aluminum 2.0 

Iron 0.50

a 1 pCi/m 3 = 0.037 Bq/m3 .

A.3-24
Table A.3.14 Exposure Conditions for a Medium-Sized Warehouse

Absorbers 

Exposure Concentration Exposure 

Exposure Time of 3H in Air Distance Material Thickness 

Event (h) (pCi/m3)a (cm) (cm) 

2 PRIMARY LOADERS AND UNLOADERS 

Handle cargo 0.033 0.0088 30 Wood 1.0 

In trailer 2.0 0.0088 120 Wood 60 

In trailer bay 4.0 0.00073 460 Wood 120 

Aluminum 1.0 

4 OTHER LOADERS AND UNLOADERS (2 PER BAY) 

1 bay away 6.0 0.00019 910 Wood 230 

Aluminum 

2 bays away 6.0 0.000048 1,400 Wood 340 

Aluminum 1.0 

3 FORKLIFT OPERATORS 

Move cargo 0.050 0.00076 120 Wood 1.0 

Iron 0.16 

Other work 40 0.00029 1,000 Wood 250 

Aluminum 10 

Iron 1.0 

6 STOREROOM CLERKS 

Handle cargo 0.033 0.00029 30 Wood 1.0 

Near cargo 8.0 0.00029 120 Wood 30 

Other work 32 0.00029 610 Wood 15 

Aluminum 2.0 

I ron 0.50

a 1 pCi/m 3 = 0.037 Bq/m 3.

A.3-25
Table A.3.15 Exposure Conditions for a Large Retail Store

Absorbers 

Exposure Concentration Exposure 

Exposure Time of 3H in Air Distance Material Thickness 

Event (h) (pCi/m 3)a (cm) (cm) 

2 PRIMARY CLERKS 

Handle 3.3 0.000088 30 Wood 1.0 

product 

Near stored 250 0.000088 460 Wood 3.0 

product 

Near display 250 0.000088 310 Wood 75 

Aluminum 1.0 

Other activities 1,500 0.000088 910 Wood 230 

Aluminum 1.0 

10 OTHER CLERKS 

Near stored 100 0.000088 460 Wood 3.0 

product 

Near display 130 0.000088 460 Wood 75 

Aluminum 1.0 

Other activities 1,800 0.000088 910 Wood 230 

Aluminum 1.0 

100,000 CUSTOMERS 

Examining 0.083 0.000088 30 Wood 1.0 

product 

Near display 5.0 0.000088 310 Wood 75 

Aluminum 1.0 

General 55 0.000088 910 Wood 230 

shopping Aluminum 1.0 

a 1 pCi/m 3 = 0.037 Bq/m 3.

A.3-26
Table A.3.16 Exposure Conditions for a Small Retail Store

Absorbers 

Exposure Concentration Exposure 

Exposure Time of 3H in Air Distance Material Thickness 

Event (h) (pCi/m3 )a (cm) (cm) 

1 PRIMARY CLERK 

Handle product 3.3 0.00039 30 Wood 1.0 

Near product 250 0.00039 Wood 3.0 

310 Aluminum 1.0 

Other activities 1,800 0.00039 460 Wood 120 

Aluminum 1.0 

2 OTHER CLERKS 

Near display 100 0.00039 310 Wood 3.0 

Aluminum 1.0 

Other activities 1,900 0.00039 460 Wood 120 

Aluminum 1.0 

10,000 CUSTOMERS 

Examining 0.083 0.00039 30 Wood 1.0 

product 

Near display 9.9 0.00039 150 Wood 50 

Aluminum 

1.0 

General 50 0.00039 460 Wood 120 

shopping Aluminum 1.0

a 1 pCi/m 3 = 0.037 Bq/m 3.

A.3-27
Table A.3.17 Exposure Conditions for an Air-Freight Terminal

Absorbers 

Exposure Concentration Exposure Exposure Time of 3

H in Air Distance Material Thickness 

Event (h) (pCi/m3)a (cm) (cm) 

2 PRIMARY LOADERS AND UNLOADERS 

Handle cargo 0.017 0.030 30 Wood 1.0 

In truck 0.5 0.030 90 Wood 30 

In trailer bay 2.0 0.00073 310 Wood 120 

Aluminum 1.0 

4 OTHER LOADERS AND UNLOADERS (2 PER BAY) 

1 bay away 2.0 0.00073 910 Wood 230 

Aluminum 1.0 

2 bays away 2.0 0.00073 1,400 Wood 340 

Aluminum 

3 FORKLIFT OPERATORS 

Move cargo 0.050 0.000078 120 Wood 1.0 

Iron 0.16 

Other work 2.0 0.00010 1,000 Wood 250 

Aluminum 10 

Iron 1.0 

6 SORTERS 

Handle 0.017 0.00010 30 Wood 1.0 

packages 

Near 2.0 0.00010 310 Wood 30 

packages 

12 FREIGHT CLERKS 

Fill container 0.017 0.00085 310 Wood 1.0 

Near 1.0 0.000065 120 Lucite 1.5 

container Wood 60 

Other work 1.0 0.000033 460 Lucite 1.5 

Wood 120 

Aluminum 1.0

A.3-28
Table A.3.17 Exposure Conditions for an Air-Freight Terminal (continued) 

Absorbers 

Exposure Concentration Exposure 

Exposure Time of 1H in Air Distance Material Thickness 

Event (h) (pCi/m3

)a (cm) (cm) 

12 PLANE LOADERS 

Move 0.083 0.0000066 90 Lucite 1.5 

container Wood 1.0 

Aluminum 0.16 

Load 0.25 0.00036 310 Lucite 1.5 

container Wood 250 

Aluminum 10 

Iron 1.0 

Other work 0.75 0.00000019 410 Lucite 1.5 

Wood 250 

Aluminum 10 

Iron 1.0 

a 1 pCi/m 3= 0.037 Bq/m 3.

A.3-29
Table A.3.18 Exposure Conditions in an Airplane

Absorbers 

Exposure Concentration Exposure 

Exposure Time of 3H in Air Distance Material Thickness 

Event (h) (pCi/m 3

)a (cm) (cm) 

3 FLIGHT CREW MEMBERS 

In cockpit 2.5 0.0011 1,100 Lucite 1.5 

Aluminum 2.0 

Wood 20 

3 FLIGHT ATTENDANTS 

Either hold 0.50 0.0011 Lucite 1.5 

Aluminum 1.0 

90 Wood 10 

Other 2.0 0.0011 1,000 Lucite 1.5 

locations Aluminum 3.0 

Wood 100 

1,200 Aluminum 3.0 

Wood 100 

1 MAXIMALLY EXPOSED PASSENGER 

Over hold 2.5 0.0011 Lucite 1.5 

Aluminum 1.0 

90 Wood 10 

120 AVERAGE PASSENGERS 

Both holds 2.5 0.0011 Lucite 1.5 

200 Aluminum 2.0 

Wood 20 

a 1 pCi/m3 = 0.037 Bq/m 3.

A.3-30
A.4 OTHER MODELING METHODOLOGIES 

A.4.1 Estimation of External Photon Dose Due to Bremsstrahlung Produced 

by Low-Energy Electrons in Beta Decay 

A.4.1.1 Introduction 

In some of the dose assessments presented in this report, estimates of external dose from 

exposure to photons emitted by radionuclides were obtained using CONDOS II (Computer 

Codes, O'Donnell et al., 1981). For radionuclides that undergo beta decay, external doses 

calculated using the CONDOS II code include a contribution due to bremsstrahlung, which is the 

continuous spectrum of photons that results when an emitted electron (beta particle) is 

decelerated by scattering in matter. This appendix presents an evaluation of the validity of 

CONDOS II in estimating external dose due to bremsstrahlung, and it describes the method used 

to adjust the dose estimates given by CONDOS II to obtain more realistic results. 

In the continuous spectrum of bremsstrahlung resulting from beta decay of radionuclides, the 

number of photons per unit energy decreases rapidly with increasing energy between zero (0) 

and a maximum energy equal to the endpoint energy of the continuous spectrum of beta 

particles, and the energies of most of the photons are only a small fraction of the beta endpoint 

energy (Evans, 1955). Thus, for example, when the endpoint energy of the beta spectrum is a 

few hundred keV, the energies of most of the photons in the spectrum of bremsstrahlung are a 

few tens of keV or less. At these low energies, estimates of external dose are the most 

uncertain, especially when exposed individuals are assumed to be located in close proximity to 

small sources. The difficulty in estimating external dose in these cases results, in part, from the 

substantial variations in the energy and angular distributions of the radiation field over the body 

surface and the rapid decrease in dose with decreasing photon energy and with increasing depth 

in the body. 

During routine use of some products or materials containing radionuclides that emit only beta 

particles, external exposure is the only credible pathway and the calculated dose is due entirely 

to bremsstrahlung. However, use of CONDOS II to evaluate doses raises the concern that it may 

significantly overestimate the external dose due to bremsstrahlung, especially for radionuclides 

that emit only low-energy beta particles. 

The evaluation of CONDOS II in this appendix focuses on calculations of external dose from 

exposure to low-energy beta-emitting radionuclides. Particular radionuclides of concern include 14C, 63Ni, and 1

47Pm, which have beta endpoint energies of 156 keV, 66 keV, and 225 keV, 

respectively (Kocher, 1981). For 3H, the dose due to bremsstrahlung is assumed to be zero (0) 

as explained in Section A.4.1.4. As noted above, it is for such low-energy beta-emitting 

radionuclides that the spectrum of bremsstrahlung is dominated by very low-energy photons and 

estimates of external dose are the most uncertain. 

The validity of CONDOS II in estimating external dose due to bremsstrahlung was evaluated by 

comparing estimates of dose from exposure to beta-emitting radionuclides obtained using the 

code with estimates given in Federal Guidance Report No. 12 (EPA-402-R-93-081). Based on 

this evaluation, a simple adjustment factor was developed and applied to calculations using 

CONDOS II to estimate external doses due to bremsstrahlung in assessments of products or

A.4-1
materials containing low-energy beta-emitting radionuclides. Other approaches to obtaining 

more realistic estimates of the external dose due to bremsstrahlung were also considered. 

However, they were not adopted for the reasons discussed below. 

A.4.1.2 Calculation of Bremsstrahlung Dose in CONDOS II Code 

CONDOS II uses the method developed by Evans (1955) to determine the external dose due to 

bremsstrahlung. Specifically, the external dose is calculated by using an approximate representation of the continuous spectrum of photons produced by the scattering of beta particles emitted by radionuclides. In this approximation, the number of photons of 

energy E. per MeV per beta emission, denoted by d(EY)/dE., is given by 

f WO (W - W,)N(W dW 

dl(E) _ "1.02kZas (1) dEV N(Vv dW 

In this equation, kis a constant equal to 7x10 4 MeV-, Zabs is the atomic number of the 

absorbing material in which the bremsstrahlung is produced, Wis the total energy of a particular beta particle in units of its rest-mass energy of 0.51 MeV given by 

W = E +1, (2) 0.51(2 

where Eis the kinetic energy of the beta particle in MeV, W0 is the value of Wcorresponding to the endpoint energy of the beta spectrum in the decay of the radionuclide, W, is the value of W 

corresponding to the photon energy Fý, and N(W) dWis the continuous spectrum of electrons 

produced in beta decay represented by 

N(W) dW = const (W 2 - 1)1

/

2(W0 - W) 2W dW. (3) 

In this approximation, the total number of bremsstrahlung photons per beta decay depends on 

the atomic number of the absorbing material, Zabs, but the shape of the spectrum of photons does not depend on the absorbing material. 

There may be considerable uncertainty in the approximations used to calculate the spectrum of 

bremsstrahlung photons described above. The constant of kin Equation 1 appears to be uncertain by about a factor of 2 (Evans, 1955). An uncertainty in the value of k results in the 

same uncertainty in the total number of bremsstrahlung photons, but there would be no effect on their energy distribution. The representation of the continuous spectrum of beta particles given by Equation (3) is an approximation that is exact only for a radionuclide with Z= 0 (Evans, 1955). The uncertainty in this approximation affects the energy distribution of bremsstrahlung photons.

A.4-2
In implementing Equations (1) through (3) in CONDOS II, the range of photon energies between 

zero (0) and the endpoint energy of the beta spectrum is divided into a number of energy 

intervals (Table A.1 of O'Donnell et al. (Computer Codes, 1981)), and the photon intensities and 

resulting external doses are calculated at each of the discrete energies used to represent these 

intervals. In the lowest energy interval, which includes the greatest number of bremsstrahlung 

photons and is of greatest interest in evaluating the validity of calculations using CONDOS II, all 

photons with energies between zero (0) and 12 keV are assumed have an energy of 10 keV. 

A.4.1.3 Comparison of CONDOS II Code With Federal Guidance Report 

Federal Guidance Report No. 12 (EPA-402-R-93-081) contains current Federal guidance on 

external dose coefficients (i.e., external dose-equivalent rates per unit concentration) for 

exposure to radionuclides in air, water, and soil. These dose coefficients were obtained using sophisticated numerical methods that represent the current state-of-the-art in calculations of 

external dose due to exposure to radionuclides dispersed in the environment. 

For beta-emitting radionuclides, the external dose coefficients given in Federal Guidance Report 

No. 12 include a contribution due to bremsstrahlung. In contrast to the approximate methods 

used in CONDOS II, the dose due to bremsstrahlung included in Federal Guidance Report 

No. 12 was calculated using realistic representations of the energy spectrum of bremsstrahlung 

for a particular energy of an emitted beta particle and the energy spectrum of electrons in beta 

decay of a particular radionuclide. These representations were obtained using sophisticated 

numerical methods. 

Another important difference between CONDOS II and the methods used in Federal Guidance 

Report No. 12 is in the approach to considering the lowest energy bremsstrahlung photons in any 

beta decay and the bremsstrahlung resulting from beta spectra with low endpoint energies. In 

particular, the following two assumptions were used in Federal Guidance Report No. 12 in 

calculating external dose due to bremsstrahlung: 

The external dose due to bremsstrahlung was assumed to be zero (0) for all photon 

energies less than 10 keV. 

The external dose due to bremsstrahlung was assumed to be zero (0) for any beta 

endpoint energy less than 100 keV. 

The second assumption is based on the first because, for beta endpoint energies less than 

100 keV, most of the bremsstrahlung photons have energies less than 10 keV. 

Thus, the dose calculations in Federal Guidance Report No. 12 ignore the lowest energy 

bremsstrahlung photons by using energy cutoffs in the spectrum of photons and in the endpoint 

energy of beta spectra, but no such cutoffs are used in CONDOS I1. This difference is 

particularly important when nearly all photons have very low energies, as is the case for low beta 

endpoint energies. The primary justification for use of energy cutoffs in Federal Guidance 

Report No. 12 is that the external dose from the lowest energy bremsstrahlung photons should 

be insignificant in any credible scenarios for exposure to radionuclides in the environment, when 

all radionuclides and exposure pathways are taken into account.

A.4-3
In this evaluation, external doses calculated using CONDOS II are compared with the dose 

coefficients tabulated in Federal Guidance Report No. 12 (EPA-402-R-93-081). The 

calculations were performed for different beta-emitting radionuclides having a wide range of beta 

endpoint energies. The source was assumed to be a uniformly contaminated ground surface, 

which can be represented as a very large disk in CONDOS II (Computer Codes, O'Donnell et al., 1981), and the dose was calculated at 1 meter above the ground. Of the source distributions 

included in Federal Guidance Report No. 12, a contaminated ground surface is the most 

appropriate for this comparison because the sources of concern in assessments of products or 

materials containing beta-emitting radionuclides are usually assumed to be point sources and a 

plane source can be regarded as an infinite array of unshielded point sources at varying 

distances from a receptor. 

The comparison of external dose coefficients for a contaminated ground surface calculated using CONDOS II with the corresponding dose coefficients given in Federal Guidance Report No. 12 is 

shown in Table A.4.1. Excluding the radionuclides with beta endpoint energies less than 

100 keV, the results of this comparison may be summarized as follows: 

For all radionuclides, the dose calculated using CONDOS II is higher than the 

corresponding value given in Federal Guidance Report No. 12. 

The degree of overestimation in the doses calculated using CONDOS II is the highest when the beta endpoint energy is close to the cutoff of 100 keV used in Federal Guidance 

Report No. 12. The substantial differences in these cases presumably reflect, at least in 

part, the use of a photon energy cutoff of 10 keV in Federal Guidance Report No. 12, in 

contrast to the inclusion of all low-energy photons in CONDOS II, because the spectrum of bremsstrahlung is dominated by the lowest energy photons. However, other 

differences between the approximate methods used in CONDOS II and the more rigorous 

methods used in Federal Guidance Report No. 12 also could be important. 

The difference between the two dose estimates decreases as the beta endpoint energy 

increases. This result presumably reflects the decreasing importance of the lowest 

energy photons as the maximum photon energy in the spectrum of bremsstrahlung 

increases. 

The difference in the dose estimates for 9°Sr/90Y is only 25%. This comparison suggests that the approximate methods used in CONDOS !1 give reasonable results for 

radionuclides having the highest beta endpoint energies (i.e., when there are significant intensities of higher energy photons in the spectrum of bremsstrahlung). 

For radionuclides with beta endpoint energies less than 100 keV, the comparisons at higher endpoint energies suggest that CONDOS II substantially overestimates the dose due to 

bremsstrahlung. The degree of overestimation cannot be determined, but it probably is greater than that found for radionuclides with beta endpoint energies slightly above 100 keV. 

A.4.1.4 Approach to Estimating Dose Due to Bremsstrahlung 

As noted in Section A.4.1.1, the particular beta-emitting radionuclides of concern to this report in 

regard to estimating external dose due to bremsstrahlung include 14C, 63Ni, and 147

Pm. These 

radionuclides have relatively low beta endpoint energies and, as indicated in Table A.4.1,

A.4-4
calculations using CONDOS II probably overestimate the external dose due to bremsstrahlung 

by a substantial amount, i.e., by more than a factor of 10. For 3H, the dose due to 

bremsstrahlung is assumed to be zero (0) as explained below. 

A simple approach that should provide more realistic estimates of external dose from exposure to 

"14C and '

47Pm is to reduce the doses calculated using CONDOS II based directly on the ratios of 

doses given in Table A.4.1. In particular, these results suggest that calculated doses should be 

reduced by a factor of 20 for 14C and a factor of 15 for '

47Pm. Such a reduction should be 

reasonable even though the comparison in Table A.4.1 applies to exposure to a contaminated 

ground plane, whereas exposure to a point source is the usual assumption in the assessments of 

products or materials containing these radionuclides. 

A similar reduction in calculated doses from exposure to 63Ni based on the results in Table A.4.1 

is more uncertain. One option would be to ignore external doses for "

3

Ni calculated using 

CONDOS II (i.e., a dose of zero (0) could be assumed in all assessments). This option would be 

consistent with current Federal guidance (EPA-402-R-93-081). A second option would be to 

reduce calculated doses for 63Ni by a factor of 20, based on the reduction factor for "4C obtained 

from Table A.4.1 and the observation that the discrepancy between doses calculated using the 

CONDOS II code and results given in Federal Guidance Report No. 12 increases with 

decreasing beta endpoint energy. With this assumption, the dose estimates for 63Ni should be 

more realistic but still conservative. The second option is used in this report, primarily because 

the external dose due to bremsstrahlung resulting from decay of 63Ni is important only in 

assessments where no other exposure pathways are assumed to occur. 

The beta endpoint energy for 3H, which also is included in Table A.4.1, is very low. Since the 

energies of nearly all bremsstrahlung photons are less than 1 keV, the most reasonable option is 

to assume that the dose due to bremsstrahlung is zero (0), which again would be consistent with 

current Federal guidance. This assumption also can be justified on the grounds that all 

assessments of products or materials containing 3H assume that some release of activity from a 

source occurs, and the resulting dose due to inhalation or absorption through the skin generally 

would be much more important than the external dose due to bremsstrahlung. 

A.4.1.5 Alternatives to Estimating Dose Due to Bremsstrahlung 

For the low-energy beta-emitting radionuclides listed in Table A.4.1, the alternative of replacing 

dose estimates obtained using CONDOS II with estimates obtained using more rigorous and 

sophisticated methods, such as those described in Federal Guidance Report No. 12 

(EPA-402-R-93-081), could be considered. Such calculations would require the use of 

complex computer codes. 

In considering whether the use of more rigorous and sophisticated calculations in estimating 

external dose from exposure to low-energy beta-emitting radionuclides is justified, the limitations 

of any such calculations in regard to obtaining realistic estimates of dose due to bremsstrahlung 

should be recognized. These limitations result from two factors: (1) the primary importance of 

very low-energy photons in the spectrum of bremsstrahlung for these radionuclides, and (2) the 

assumption of idealized exposure conditions in any dose assessment (e.g., exposure at a fixed 

distance from a point source shielded only by air or a simple configuration of another material). 

Even the most sophisticated calculations would not provide an accurate accounting of the 

significant scattering and absorption of the lowest energy photons in materials used in the source

A.4-5
mounting, other materials located close to the source (e.g., in an instrument housing), and 

clothing worn by an exposed individual. Furthermore, it is very difficult to accurately account for 

the substantial variations in the energy and angular distributions of low-energy photons over the 

body surface of an exposed individual located close to a source, with the result that there would 

be considerable uncertainty in estimated doses at different depths in the body. Finally, the 

energy and angular distributions of low-energy photons at the body surface, and thus the dose at 

different depths in the body, would be quite sensitive to the assumed distance of an exposed 

individual from a source. 

Based on these considerations, calculations of external dose due to bremsstrahlung performed using complex computer codes are unlikely to produce realistic results for low-energy beta

emitting radionuclides. Therefore, this alternative is unlikely to have a substantial benefit 

compared with an approach of applying simple reduction factors to dose estimates obtained 

using CONDOS I1. 

A.4.1.6 Conclusion 

A simple approach to addressing the likely overestimates of external dose due to bremsstrahlung obtained using CONDOS II is adopted for use in this report. In this approach, simple radionuclide-specific reduction factors are applied to calculated doses for the low-energy beta 

emitters of concern. These reduction factors are given in Table A.4.2. The reduction factors for 14

C and 14 7

Pm are obtained directly from a comparison of doses calculated using CONDOS II 

with the results given in Federal Guidance Report No. 12 for the same source configuration (see Table A.4.1). For 63

Ni, which has a beta endpoint energy less than 100 keV, the dose due to 

bremsstrahlung is assumed to be zero (0) in Federal Guidance Report No. 12. In this case, dose 

estimates obtained using CONDOS II are adjusted using the assumed reduction factor for 1

4C. 

The dose estimate obtained using this reduction factor should be more realistic but still 

conservative. Finally, the external dose due to exposure to 3

H is assumed to be zero (0) in all 

cases, due to the very low energies of all bremsstrahlung photons and the assumption in all 

assessments that pathways of internal exposure would occur. 

A.4.2 Generic Dose Modeling for Sources in Close Proximity to the Body 

Several of the exposure scenarios involve the placement of the item/source in close proximity of 

the body/skin surface. For example, several misuse scenarios call for the user to carry the 

exempt item in the shirt pocket or coveralls over an extended time period. Other routine 

exposure scenarios involve the item/source being in contact with the skin, such as for a watch 

dial using tritiated paint. 

The calculations of the effective dose equivalent (EDE) and the localized skin dose from a small 

source in close proximity to the body is complicated by several factors. Radiation fluence will 

decrease rapidly with increased distance from the source; electron attenuation by clothing becomes critical for skin dose calculations. For the calculation of EDE, the body will not be 

exposed to a uniform radiation field. Therefore, conventional methods for correlating fluence to 

effective dose equivalent are not appropriate. Use of the 1 cm deep-dose equivalent, as 

specified in 10 CFR 20, or the fluence-to-dose conversion factors of the International 

Commission on Radiation Units and Measurements (ICRU) (ICRU 47) would yield unrealistic 

estimates of the EDE. For calculation of the skin dose, placing the source in a pocket of clothing

A.4-6
will reduce the electron energy level and fluence. Assumptions regarding thickness and density of clothing can have a significant effect on resultant skin dose calculations. 

For the calculation of the localized skin dose, VARSKIN MOD2 (Computer Codes, Durham, 

1992) provides a method for performing the calculations. Generic assumptions regarding distances and shielding materials (cloth covering) are needed in the calculations for consistency. 

However, for the EDE, attempting to calculate doses to the different internal organs and applying weighting factors becomes quite involved, requiring analytical modeling techniques that do not 

lend themselves to simple applications like these. A simple method is needed to estimate the 

EDE based on assumptions concerning location and tissue depth. Such a method would allow 

the use of readily available radiation shielding approximation methods, such as MicroShield 

(Computer Codes, Grove Engineering, 1996), for estimating the EDE. 

An in-depth evaluation of the EDE for photon radiation sources external to the body was 

performed by Reece, et. al., for the Electric Power Research Institute (EPRI) (Reece, 1993). This evaluation presented calculations of the EDE for point sources located on the external body surface. Mathematical models of the adult male and female body were coupled with Monte Carlo 

modeling of photon source and transport for the calculations. Doses to internal organs were 

calculated for specified source-organ geometries; organ weighting factors were applied; and the 

EDE was calculated by summing the weighted organ doses. 

Using this method, the EDE was calculated with the point source at varying locations on the torso 

of the body. The position of the source was incrementally increased circumferentially around the 

body and vertically up from the base of the torso to 61 cm. The EDE was calculated for a total of 

52 locations for a point source located on the torso of an adult male and adult female. For the 

male, the highest EDE was for the source located at the front part of the torso at a 6 cm height above the base, which resulted primarily from the dose to the gonads. For the female, the 

highest calculated EDE was with the source located in front at a height of 61 cm, which resulted 

primarily from the dose to the breast. 

The methods employed in the EPRI study were very detailed, requiring extensive modeling and 

computer application. For this study, a simpler technique was needed. The approach selected 

was to establish an effective tissue depth that would allow the use of the point kernel radiation 

shielding code MicroShield (Computer Codes, Grove Engineering, 1996) for estimating an EDE 

for sources in close proximity to the body. Using MicroShield, the deep dose equivalent at a 

10 cm depth in tissue was compared with the results from the EPRI study. The 10 cm depth was 

considered representative of the typical depth to the radiosensitive organs, as used for 

calculating the EDE. 

As evaluated in the EPRI study, three different photon energies were evaluated - 0.08, 0.3 and 

1.0 MeV. For 0.08 MeV photons, calculations using MicroShield (Computer Codes, Grove 

Engineering, 1996) for the deep dose equivalent were a factor of five greater than the maximum 

calculated EDE in the EPRI study for a point source in contact with the torso of the body. For 

0.3 MeV photons, MicroShield calculations were a factor of 3 greater and a factor of 2 greater for 

1.0 MeV photons. Based on these comparisons, it was concluded that using MicroShield and 

calculating the deep dose equivalent at a 10 cm tissue depth provided a reasonably conservative 

approach for estimating the EDE for sources in close proximity to the body.

A.4-7
In applying this approximation method for the different exemptions, the source cannot be 

expected to be in direct contact with the body on a continuous basis. Therefore, for most 

applications, a 1 cm separation has been assumed from the source to the skin of the body. 

Additionally for the evaluation of the electron dose component, attenuation by typical clothing has 

been included. The source is assumed to be placed in a clothing pocket having a thickness of 

0.07 cm and a density of 0.4 g/cm 3, yielding an effective thickness of 0.028 g/cm' which is a 

value typical for cotton coveralls (Martz et al., 1986). 

Thus, for the purposes of generic modeling the following assumptions have been made: 

source located in a pocket of clothing, typically 1 cm from the body surface, 

clothing has a thickness of 0.07 cm and a density of 0.4 g/cm3, yielding an effective 

thickness of 0.028 g/cm2, and 

effective dose equivalent is estimated by calculation of the deep dose equivalent using 

MicroShield at a tissue depth of 10 cm, 21 

21 MicroShield calculations performed with a total distance of 11 cm, comprised of 1 cm of air 

(source distance from body), 9 cm of tissue (shielding), and 1 cm of air (recognizing that the 

deep dose equivalent fluence-to-dose factors are based on a 1 cm depth).

A.4-8
Table A.4.1 Comparison of External Doses Due to Bremsstrahlung for Exposure 

to Contaminated Ground Surface Calculated Using the CONDOS II Code 

With External Dose Coefficients Given in Federal Guidance Report No. 12 

External Dose Coefficient 

(Sv/s per Bq/m2) 

Radionuclidea CONDOS II Federal Guidance Report No. 12b Ratioc 

3H (19 keV) 3.7x10-21 0d 

'Ni (66 keV) 5.6x1 0-21 d 

14

C (156 keV) 3.4x10-19 1.6x10-20 21 

147 Pm (225 keV) 5.1x10-19 3.4x10-20 15 

"3Cl (710 keV) 2.1x10- 18 6.7x10- 19 4.6 

"09

Sr/°Y (2,284 keV)e 7.Oxl -18 5.6xl 0-18 1.25 

a Entry in parentheses is the endpoint energy of the continuous spectrum of electrons in beta 

decay (Kocher, 1981). Radionuclides are listed in order of increasing beta endpoint energy. b Dose coefficients given in Table 111.3 of Federal Guidance Report No. 12 

(EPA-402-R-93-081). 

c Value obtained using CONDOS II divided by the value in Federal Guidance Report No. 12. 

d Dose is assumed to be zero (0) when beta endpoint energy is less than 100 keV. 

e Beta endpoint energy is the value for the short-lived decay product 90Y, which is assumed to be 

in activity equilibrium with 9

"Sr; beta endpoint energy for 9°Sr is 546 keV.

A.4-9
Table A.4.2 Dose Reduction Factors Applied to Estimates 

of External Dose Due to Bremsstrahlung Obtained 

Using the CONDOS II Code a

Radionuclide Dose Reduction Factor 

3H b 

14c 20 

63

Ni 20 

147pm 15 

a For the other beta-emitting radionuclides of concern to this report, correction of external doses 

estimated using CONDOS II is unimportant, either because the correction factor is small for high

energy beta emitters or because the radionuclide also emits gamma rays or X-rays, or pathways 

of internal exposure are assumed to occur. 

b Dose due to bremsstrahlung is assumed to be zero (0), because the energies of the 

bremsstrahlung photons are very low and pathways of internal exposure also are assumed to 

occur.

A.4-1 0
APPENDIX B
GLOSSARY

bastnasite-a light lanthanide (Ln) fluoride carbonate that occurs in an unusual type of magma

derived deposit in which the Ln-elements have been enhanced. 

beta backscatter/transmission devices--devices that use particles from a variety of sources to 

measure the thickness or density of thin films and thin coatings on other materials. 

byproduct material-any radioactive material (except special nuclear material) yielded in or made 

radioactive by exposure to radiation during the process of producing or utilizing special nuclear 

material. 

calibration source-a source of a known purity and activity that is used to determine the variation 

in accuracy of a measuring instrument and to ascertain necessary correction factors. 

chemical detectors-devices used to monitor for harmful or toxic gases and a variety of vapors. 

collective dose-the sum of the individual doses received in a given period of time by a specified 

population from exposure to a specified source of radiation. 

decommission-to remove safely from service and reduce residual radioactivity to a level that 

permits release of the property for unrestricted use and termination of license. 

depleted uranium-the source material uranium in which the isotope uranium-235 is less than 

0.711 percent by weight of the total uranium present. Depleted uranium does not include special 

nuclear material. 

disposal-isolation of radioactive wastes from the biosphere inhabited by man and containing his 

food chains by emplacement in a land disposal facility. 

dose-a general term denoting the quantity of radiation or energy absorbed per mass of tissue. 

For special purposes it must be appropriately qualified. If qualified, it refers to absorbed dose. 

dose commitment-the total radiation dose to a part of the body that will result from retention in 

the body of radioactive material. For purposes of estimating the dose commitment, it is assumed 

that from the time of intake the period of exposure to retained material will not exceed 50 years. 

effective dose-the sum of the weighted equivalent doses in all the tissues and organs of the 

body. 

electron capture detectors for gas chromatographs-devices used to identify molecules in the 

effluent stream from gas chromatographs. 

exempt concentrations--some generally licensed items contain only small quantities of 

byproduct material, and these items are potential candidates for exemption from licensing 

requirements.

B-1
external dose-that portion of the dose equivalent received from radiation sources outside the 

body. 

fission-the splitting of a nucleus into at least two other nuclei and the release of a relatively large 

amount of energy. Two or three neutrons are usually released during this type of transformation. 

general license-the possession and use of specified quantities of certain radionuclides without 

the need for applications or issuance of licensing documents to the particular persons using the 

radioactive materials. 

incandescent gas mantles-mantles containing thorium that are available in a variety of designs and sizes, each intended to fit into one of the many different lighting devices in use, such as 

camping lanterns, recreational vehicle lights, and outdoor gaslights. 

incinerator workers-individuals who engage in sweeping or other cleanup activities while 

located at the edge of a partially enclosed tipping area where garbage trucks unload waste at the 

facility. 

ion generating tube-devices designed for ionization of air that contains, as a sealed source or 

sources, byproduct material consisting of a total of not more than 19 megabecquerel 

(500 microcurie) of polonium-21 0 or 1.9 gigabecquerel (50 millicurie) of tritium per device. 

internal dose-that portion of the dose equivalent received from radioactive material taken into 

the body. 

landfill workers-individuals who are located on top of the waste pile during operations and 

perform such tasks as dumping of waste, grading of the waste following dumping, and covering 

of the waste at periodic intervals. 

liquid scintillation counters-devices that measure light emitted by a scintillator medium in which 

radioactive materials are intimately dispersed and estimate the concentration of the radioactive 

material from the light intensity. 

maximum permissible concentration (MPC)-an acceptable upper limit for the concentration of a 

specified radionuclide in a material taken into the body, below which continuous exposure to the 

material will result in acceptable health risks to the specified population involved. 

monazite-a rare earth phosphate, one of the most abundant rare earth minerals. 

ore-a natural mineral compound of the elements of which at least one is a metal. 

piezoelectric ceramics-used in many different shapes and sizes in consumer products that 

require an electromechanical coupling device. Produced when pressure is applied to certain 

classes of crystalline materials, where the crystalline structure produces a voltage proportional to 

the pressure. Such consumer products include pacemakers, electronic telephone ringers, 

microphones, patio grills, and games and toys. 

rad-the unit of absorbed dose equal to 0.01 joule (J)/kg or 0.01 gray in any medium.

B-2
radioactive tracers-substances used to label specific atoms, molecules, living organisms, or 

other entities. The tracer may be used to study the kinetics of exchange, distribution, 

metabolism, turnover, conversion, and excretion of the labeled compound. 

radiation monitoring--the measurement of radiation levels, concentrations, surface area 

concentrations, or quantities of radioactive material and the use of the results of these 

measurements to evaluate potential exposures and doses. 

rem--a special unit of dose equivalent. The dose equivalent in rem is numerically equal to the 

absorbed dose in rad multiplied by the quality factor, the distribution factor, and any other 

necessary modifying factors. 

sealed source-any licensed material that is encased in a capsule designed to prevent leakage 

or escape of the licensed material. 

self-luminous devices-devices that use byproduct material to create light without outside 

activation. Examples are wristwatches, leveling bubbles, automobile lock illuminators, gun 

sights, and aircraft and building exit signs. 

sievert--the unit of radiation dose equivalent that is used for radiation protection purposes for 

engineering design criteria and for legal and administrative purposes; equal to 1.0 joule (J)/kg. 

source material--uranium or thorium, or any combination thereof, in any physical or chemical 

form or ores that contain by weight one-20th of 1 percent (0.05%) or more of uranium, thorium, or 

any combination thereof. 

spark gap irradiators containing cobalt-60--irradiators designed to minimize spark delay in some 

electrically ignited commercial fuel-oil burners by generating free electrons in the spark gap. 

These free electrons are produced from beta particles emanating from the cobalt-60 plating on 

the irradiator. 

special nuclear material -plutonium, uranium-233, uranium enriched in the isotope 233 or 235, 

and any other material that the Nuclear Regulatory Commission, pursuant to the provisions of 

section 51 of the Atomic Energy Act, determines to be special nuclear material, but does not 

include source material; or any material that is artificially enriched by any of the foregoing but 

does not include source material. 

static eliminators-devices that work on the principle that static charges can be neutralized by ionized particles. These devices use up to 200 millicurie of polonium-21 0 to ionize the air where 

static charges may build up. These devices can be portable or stationary. 

thermoluminescent dosimetry readers-devices used to determine the radiation dose to an 

exposed piece of thermoluminescent material by measuring the light output as the material is 

heated. 

thorium vacuum tubes-vacuum tubes containing thoriated tungsten cathodes in many varied 

designs.

B-3
tungsten inert-gas (TIG) arc welding-a process in which an electrical arc is struck between an 

inert, gas-cooled, nonconsumable electrode (also called a welding rod), and the metal work 

pieces. Many electrodes used in TIG welding consist of tungsten wire that contains thorium 

dioxide or another metal oxide. 

uranium milling--any activity that results in the production of byproduct material. 

waste collectors-individuals who collect waste from the generating site, haul the waste to 

garbage trucks, and transport the waste to landfills. 

x-ray fluorescence analyzers-devices designed for use in nondestructive analysis to determine 

the elemental chemical composition of solid and liquid samples.

B-4
SYMBOLS AND UNITS 

Bq becquerel 

C coulomb 

Ci curie 

cm centimeter 

d day 

g gram 

Gy gray 

h hour 

kg kilogram 

km kilometer 

L liter 

m meter 

min minute 

mg milligram 

mL milliliter 

mm millimeter 

MT metric ton 

ppb parts per billion 

ppm parts per million 

p-rem person-rem 

R roentgen 

s second 

Sv sievert 

E exa 1018 

P peta 1015 

T tera 1012 

G giga 109 

M mega 106 

k kilo 103 

c centi 10-2 

m milli 10-3 Smicro 10-6 

n nano 10-9 

p pico 10-12 

f femto 10-15

B-5
INDEX

abandoned mine ........ 

aggregate ............. 

aircraft engine part ...... 

aircraft safety devices .... 

ALARA ............... 

area/process monitor .... 

asphalt ............... 

automobile glass ........ 

automobile lock illuminator 

automobile shift quadrant 

baggage .............. 

balances of precision .....

barium titanate

.......................................... 3-63,3-67 

............................... 3-23, 3-30, 3-65, 3-155 

.. .......................... ................... xxv 

.............................................. 2- 111 

........................... ..................... 1-45 

.............................................. 2-145 

............................................... 3- 23 

........................................ 3- 149,3-181 

............................................... 2-77 S.............................................. 2-98 

° 2-17,2-19,2-20, 2-22, 2-40, 2-44, 2-45, R-20, R-21, A.3-6 

............ xxx, xxxi, 2-87, 2-88, 2-90-93, 2-95, 2-106, 4-4

......................................................... 3-209

bastnasite ... xxvii, 3-33, 3-147-150, 3-152-158, 3-162, 3-170, 3-171,3-173, 3-174, 3-179, 

3-181,3-182 

benchtop instrum ent ..................................................... 2-145 

beryllium ............................................................... 3-75 

boiler slag .............................................................. 3-22 

borosilicate ............................................................ 2-195 

bottom ash ............................................................. 3-22 

bow .................................................................. 2-196 

bremsstrahlung ...... 1-11,2-50-52, 2-60, 2-63, 2-73, 2-120, 2-124, 2-127, 2-216, 2-224, 

3-195,3-219,4-17,4-19,4-20,4-31,4-34,4-37,4-42,4-54,4-55,4-59, 

4-64, A.3-2, A.4-1-6, A.4-9, A.4-10 

calibration source ................. 2-167,2-172,2-173,2-178,2-191,3-86,3-90,4-64 

casting ....................... 3-23, 3-29, 3-44, 3-50, 3-51, 3-241-243, 3-302, 3-303 

casting m aterial .......................................................... 3-23 

ceramic or alumina bead ................................................. *2-161 

ceramic tableware . . xxv, xxvii, xxx, xxxi, 3-191, 3-192, 3-194-201, 3-203, 3-204, 3-207, R-1 7 

cerium ... xxvii, 2-182, 3-75, 3-103, 3-147-151, 3-153, 3-155-163, 3-169-171, 3-173, 3-174,

chemical detector ............... 

chlorinator bed ................. 

class Y material ................. 

cloisonne jewelry ................ 

coal ash ....................... 

communication equipment ........ 

com post ...................... 

contaminated steel .............. 

crushed base .................. 

cyanoacrylate .................. 

decolorization of glass ............ 

dental porcelain ................. 

dental product .................. 

depleted uranium . xxvi, 3-2, 3-3, 3

3-260, 3-269, 3-270, 3-

3-176, 3-180-182, R-11 

xxiii, 2-218, 2-224, 2-226-228, 2-234, 2-236, 2-238 

..................................... 3-50,3-51 

................................... 3-108,3-113 

......................................... 3-232 

............ 3-20, 3-22-24, 3-30, 3-31, 3-41, 3-52-56 

......................................... 2- 119 

.......................................... 3-23 

........ 2-19, 2-20, 2-24, 2-25,2-27, 2-30, 2-33,2-48 

.......................................... 3- 23 

......................................... 2-196 

......................................... 3-149 

...................... 3-20, R-1, R-15, R-19, R-22 

.......................................... 3- 24 

5, 3-8, 3-19, 3-20, 3-40, 3-42, 3-43, 3-249-251, 3-258, 

-274, 3-277, 3-279, 3-281 -283, R-4, R-7, R-12, A.1-17,

I-1
A.1-19, A.1-22, A.2-33, A.2-35, A.2-37, A.2-39, A.2-41, A.2-44, A.2-46, 

A.2-48, A.2-50, A.2-52, A.2-54, A.2-56, A.2-58, A.2-60, A.2-61, B-1 

elastom eric ............................................................ 2-195 

electric lam p ........................................................... 3-126 

electromechanical coupling device ...................................... 3-209, B-2 

electron tube .................................. 2-119-121,2-123-127, 2-131, 2-137 

explosive ...... 2-16,2-17,2-19, 2-20, 2-22, 2-23, 2-25,2-40, 2-44, 2-45, 4-4, 4-5, R-20, 

R-21 

fertilizer ............................................................ 3-23 

Fiestaware® ..................................................... 3- 192, 3-193 

fluorescent lamp ............................................... 3-123-127, A.1-9

fly ash ............. 

gas chromatograph ... 

gaseous tritium ...... 

gemstone .......... 

general license .... 1 

glass enamel ........ 

glass enamel frit ..... 

glaze .............. 

gun sight ........... 

hafnium ............ 

handgun ........... 

handheld instrument .. 

ilm enite ............ 

incandescent lamp ... 

insoluble oxide ...... 

internal calibration .... 

ion exchange resin ...

jewelry... 2-58, 2-64,

........................................... 3-22,3-30,3-41 ................................. xxviii, 4-30, 4-36, 4-37, 4-40 

.............................................. 2-195, A.1-9 ................. 2-17, 2-21,2-22,2-26, 2-46, 2-48, 3-163, 3-167 

-10, 2-139, 2-162, 2-194, 3-233, 4-1, 4-3, 4-15, 4-18, 4-29-31, 4-41, 

4-43, 4-53, 4-57, 4-58, 4-70, 4-71

3-27

........................... 3-231,3-232, R-18 

.......................... 3-231,3-232, R-18 .. 3-165, 3-191, 3-193-196,3-198-204,3-207, R-4 

............................ 2-207-209,2-213 

......... 2-184, 3-20, 3-22, 3-29, 3-44, 3-52, 3-53 

................. 2-196, 2-206-209, 2-213, 2-214 

....................................... 4- 16 

3-28, 3-32-35, 3-42, 3-44, 3-50, 3-51, 3-60, 3-148 

................................ 3-163,3- 167 

.......................... 2-50,3-123,3-135 

....... 2-139, 2-141-145, 2-147, 2-189,4-54, 4-57

............................................ 2-140,2-144 

2-77,2-78, 2-84, 2-85, 2-89, 2-90, 2-93, 2-98,2-103, 2-114,-2-118, 

3-231.3-232. A.1-12

lead m etaniobate ..................................... 

lead zirconate titanate ................................. 

liquid crystal display .................................. 

liquid scintillation counter ......................... ...... 

lock illum inator ................................. ..... 

magnesium-thorium alloy ................. xxx, xxxi, 3-239

m agnetron .......................................... 

m arine com pass ..................................... 

m edical glassware .................................... 

m icrowave oven ..................................... 

microwave receiver protector tube ........................ 

m icrowave tube ...................................... 

m ineral sand ................................... 3-27

.................... 3-209 

.............. 3-209,3-210 

.................... 2-195 

..................... 4-53 

........ 2-74-78,2-84, 2-85 

*243, 3-245, 3-248, 3-302-304 

...... 3-95-98, 3-100, 3-239 

.............. 2-105,2-108 

.................... 3-149 

............ 3-95-102, R-1 9 

... 2-119, 2-120, 2-125-127 

..................... R-7 . ?1--3F. ,"--2 ,3--6 9."-R .•-FRR

monazite ...... 3-6, 3-7, 3-28, 3-29, 3-33, 3-41, 3-62-68, 3-147-152, 3-154, 3-155, 3-172, 

3-177, 3-178, R-8 

narcotic ............................................................... 2-218 

navigational instrument ............................................. 2-105, 2-108 

oil and gas production and process waste ...................................... 3-24

1-2
oil w ell ................................................................ 2-162 

ophthalmic glass . xxvi, 3-20, 3-21,3-26, 3-32, 3-35, 3-38, 3-59, 3-60, 3-150, 3-153, 3-154, 

3-156, 3-159, 3-162, 3-181,3-182, R-2 

ore-buying station ................................................... 3-61,3-69 

paving ............................................................ 3-31,3-36 

personnel neutron dosimeter ......................................... 3-188, 3-190 

phosphate ore ................................................. 3-23,3-31,3-47 

phosphate slag .... xxvi, 3-20, 3-23, 3-24, 3-30-32, 3-36, 3-37, 3-47, 3-52-55, 3-57, 3-59, 

3-60

phosphogypsum ........... 

photographic camera ....... 

photographic film ........... 

piezoelectric ceramic ........ 

pointer ................... 

policy statement ........... 

porcelain teeth ............ 

pyrex .................... 

Pyrex@ .................. 

quantities of byproduct material 

radar transmit-receiver switch 

radiation measuring instrument 

radiation-resistant glass ..... 

radio broadcasting .......... 

railroad ballast ............

...................................... 3- 23, 3-47, R-5 

..................... xxv, 3-287, 3-290, 3-292-295, 3-299 

........................... xxx, xxxi, 1-3, 3-233, 4-4, 4-6 

............................. xxx, xxxi, 3-209-212, 3-214 

.................................. 2-114, 2-115, 2-118 

............................................... 1-2 

.......................................... 3-20, R-1 

.................................. 2-195, 2-196, 3-219 

............................................... 3-194 

.... xxiii, 2-2, 2-17, 2-33, 2-35-39, 2-121,2-123, 2-133-136, 

2-139,2-142, 2-147,2-163-180,2-182-188, 2-190, B-1 

.............................................. 2-120 

.............................................. 2-141 

............................................... 3-149 

............................................... 3-98 

............................................... 3- 23

rare earth ... xxiv, xxvii, xxx, xxxi, 3-2, 3-22, 3-26, 3-34, 3-50, 3-62, 3-147-159, 3-161-171, 

3-177, 3-178, 3-181, R-10, B-2 

refrigerant ............................................................. 3-165 

resin-coated sand ....................................................... 2-161 

road construction .................................................... 3-22, 3-37 

rockwool insulation ....................................................... 3-23 

rutile ................................................... 3-42,3-44,3- 50,3- 51 

sapphire .............................................................. 2-196 

scandium ....................... 2-161,2-162,2-186,3-147,3-149, 3-166, R-3, R-8 

scintillation detector ...................................................... 2-140

scrap metal .................. 3-241, 3-244, 3-245, 3-248, 3-306-308

sealed tube ............................................................ 2-195 

self-luminous ..... xxx, xxxi, 1-3, 1-6, 1-9, 2-58, 2-59, 2-73, 2-74, 2-87, 2-95, 2-96, 2-105, 

2-111, 2-193-197,2-199-203,2-205-214, 2-216, 3-166, B-3 

self-luminous paint ............................. 2-73,2-74, 2-87, 2-95,2-96, 2-111 

shipping container ................................. 3-269, 3-272, 3-273, 4-15, 4-29 

silicon semiconductor ....................... 2-16-18, 2-22-24, 2-28, 2-31, 2-34, 2-48 

smelting .................................. 3-45, 3-50, 3-217, 3-306, A.2-5, A.2-28 

smoke detector ...... 2-217, 2-219-223, 2-227, 2-229, 2-233, 3-310, 3-312, R-17, A.1-5-7 

soil stabilization .......................................................... 3-23 

solvent .................................................. 3-45, 3-156, 3-174 

sources for calibration ............................................... 2-164, 4-53 

spark gap irradiator .................................................. xxiv, 2-151 

spark gap tube ......................................................... 2-127 

1-3
special application metal .............................................. 3-29, 3-30 

sports arena ........................................... 3-138-140, 3-143, 3-145 

stabilization m aterial ...................................................... 3-23 

static eliminator ............................................... 4-4-7, 4-10, 4-13 

storage pad ............................................................. 3-67 

structural fill ............................................................. 3-22 

sulfur hexafluoride ....................................................... 2-218 

sunlamp ......................................... xxx, 3-139, 3-140, 3-143, 3-145 

tape dispenser ................................. 3-27, 3-29, 3-35, 3-36, 3-59, 3-60 

television camera ........................ xxv, 3-286, 3-289, 3-290, 3-292-295, 3-298 

thermostat dial ............................................. 2-114, 2-115, 2-118 

thorium oxide ........................... 3-23, 3-41,3-75, 3-76, 3-105, 3-240, 3-285 

timepiece ....................................... 2-49-53, 2-58-60, 2-63-66, 2-69 

tin .................... 2-12, 2-188, 3-29, 3-41,3-50,3-52, 3-53,3-148, 3-210, 3-231 

titanium .... 3-20, 3-22, 3-29, 3-42, 3-44, 3-45, 3-50-53, 3-62, 3-148, 3-164, 3-167, 4-30, 

4-31, 4-38 

topaz ......................... xxiii, 2-16, 2-17, 2-21-28, 2-34, 2-35, 2-46, 2-48, R-18 

tracer ................................................ 2-161,2-162, 3-166, B-3 

tritiated paint ........................................... 2-55, 2-59, 2-111, A.4-6 

tritium gas ....................................................... 2-105, R-17 

tungsten-thorium alloy .................................... 3- 241-244, 3-248, 3-305 

ultraviolet absorption glass ................................................ 3- 149 

unprocessed ore ........................... xxv, xxx, xxxi, 3-1, 3-61, 3-62, 3-66, 3-69 

uranium mine ................................... 3-63, 3-65, 3-67, 3-68, 3-70, R-3 

uranium oxide ................................... 3-41, 3-191, 3-210, 3-217, 3-231 

vacuum tube .................................................. 3- 100-102, R-20 

watch ...... 1-3, 2-49-53, 2-56-60, 2-62-64, 2-69, 2-77, 2-89, 2-98, 2-114, 2-127, 2-195, 

2-197-205,2-209-212, 3-100, 3-166, R-23, A.1-12, A.1-23, A.4-6 

water treatment sludge ..................................... 3-20, 3-23, 3-30, 3-31 

welding rod ................ 3-23, 3-103, 3-111,3-112, 3-114, 3-115, 3-120, 3-121, B-4 

wristwatch ................... 2-49, 2-50, 2-55-58, 2-195, 2-198-203, 2-206, 2-210-212 

xenotim e ........................................................ 3-147, 3-149 

x-ray fluorescence analyzer ............................................. xxix, 4-44 

yttrium ......................... 2-188, 3-75, 3-103, 3-147, 3-149, 3-150, 3-167, R-8 

zircon flour ................. 3-23, 3-27, 3-29, 3-32, 3-34, 3-35, 3-37, 3-38, 3-59, 3-60 

zircon sand ................................ 3-23, 3-27-29, 3-32-36, 3-38, 3-58, R-3 

zirconium .... 2-188, 3-20-23, 3-26, 3-27, 3-29, 3-32, 3-36, 3-44, 3-45, 3-52, 3-53, 3-62, 

3-103,3-148, 3-154, 3-159,3-162, 3-210, R-12, R-21

1-4
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-89) (Assigned by NRC, Add Vol., Supp., Rev., NRCM 1102. and Addendum Numbers, if any.) 3201, 3202 BIBLIOGRAPHIC DATA SHEET 

(See instructions on the reverse) NUREG-1717 

2. TITLE AND SUBTITLE 

Systematic Radiological Assessment of Exemptions for Source and Byproduct Materials 3. DATE REPORT PUBLISHED 

MONTH YEAR 

June 2001 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

S. Schneider, NRC C.R. Mattsen, NRC Final R rt 

D.C. Kocher* S.J. Cotter* epo G.D. Kerr* J.S. Bogard* 7. PERIOD COVERED (inclusive Dates) 

P.A. Scofield* J.S. Bland" 

F.R. O'Donnell* C. Wiblin** 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office orRegion, U.S. Nuclear Regulatory Commission, and mailing address; if contractor, 

provide name and mailing address.) 

Division of Risk Analysis and Applications *Oak Ridge National Laboratory **J. Stewart Bland Associates, Inc. 

Office of Nuclear Regulatory Research Oak Ridge, TN 37830 788 Sonne Drive 

U.S. Nuclear Regulatory Commission Oak Ridge, TN 37830 Annapolis, MD 21401 

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, typ "Same as above'; ff contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, 

and mailing address.) 

Same as 8. above.

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less) 

This report is an assessment of potential radiation doses associated with the current exemptions from licensing for the majority 

of Part 30 byproduct and Parts 40 and 70 source material in Title 10 of the Code of Federal Regulations (CFR). Doses were estimated for the normal life cycle of a particular produce or material, covering distribution and transport, intended or expected routine use, and disposal using ICRP 26 and 30 dose assessment methodology, which was incorporated into the current 

requirements of 10 CFR Part 20 in May 1991. In addition, assessments of potential doses due to accidents and misuse were estimated. Also presented is an assessment of potential radiological impacts associated with selected products containing byproduct material which currently may only be used under a general or specific license and may be candidates for exemption 

from licensing requirements

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT 

exemption unlimited 

dose assessment 14. SECURITY CLASSIFICATION 

Part 30 (This Page) 

Part 40 unclassified 

(This Report) 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (2-89)
Federal Recycling Program
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, DC 20555-0001 

OFFICIAL BUSINESS 

PENALTY FOR PRIVATE USE, $300