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OCR for page 371
I
Radiation Dose Assessment
Under normal operating conditions, nuclear facilities release radioac-
tive effluents in many physical and chemical forms (See Appendixes D
and E). These effluents can travel through the environment in a number
of physical pathways to expose individuals and populations surrounding
the facilities. Individuals may be exposed to radiation from immersion in
clouds of radioactive gases, inhalation of radioactive materials in the air,
ingestion of radioactive materials from contaminated foods and liquids, and
other less common pathways. Each pathway generates different patterns
of whole-body and organ exposures, often with different time courses. For
example:
• The immersion of an individual in a cloud of radioactive iodine
generates an exposure pattern characteristic of external radiation—
namely, absorbed doses delivered at various depths in tissues from
penetrating radiation (e.g., gamma rays) as well as skin exposure
due to finite-range charged particles (e.g., electrons from beta de-
cay). These doses are relatively uniform with the exception of bone
and red marrow doses, which can differ by as much as a factor of
2. These exposures persist only when the radioactive material is
present.
• Alternatively, intakes of radioactive iodine by inhalation and inges-
tion can result in exposures of individual organs, most prominently
the thyroid in the case of soluble forms of iodine. The organ doses
can vary according to biokinetic properties of radioactive iodine.
371
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372 APPENDIX I
As a result, organ-specific doses can vary significantly for different
organs.
Organ absorbed doses for these many exposure pathways have been
studied for decades and for most radionuclides. The absorbed dose to
individual organs is well established and provided in a series of reports
published by the International Commission on Radiological Protection
(ICRP). ICRP recommendations address ingestion and inhalation scenarios.
U.S. Nuclear Regulatory Commission (USNRC) licensing activities for
nuclear plants are based on the very simplistic dosimetry model reported
in ICRP Publication 2 (1959). In this model, the concept of the critical
organ is applied. The critical organ is defined as the organ, which can
include the whole body, that is expected to receive the largest radiation
dose. In contrast, current ICRP guidelines account for the exposure of
all organs and tissues. Doses from intakes of radionuclides by individuals
generally are much more accurately and comprehensively modeled under
these guidelines.
Estimating the radiation exposure to individuals in the vicinity of a
nuclear facility is a strong function of the type of facility, local conditions
such as distances from effluent release points, and of course environmental
conditions. Although there are wide variations in these conditions, esti-
mating radiation exposures reduces to knowing effluent release patterns
as a function of time, exposure pathways, and the quantity and type of
radionuclide(s) released.
Some of this required information is quite complex. For example, to
estimate radiation exposures from atmospheric release, one needs to know
radionuclide quantities, concentrations, and release locations as a function
of time, the local weather pattern also as a function of time, and any oc-
cupancy at appropriate locations surrounding the facility.
When the information discussed above is convolved with the aforemen-
tioned dosimetric models, individual and population absorbed doses can
be estimated on an individual-by-individual basis. The reliability of these
estimates will depend on the availability and quality of all the required
input data.
I.1 EXTERNAL DOSES
External doses resulting from atmospheric releases of radioactive efflu-
ents consist of three components: (1) dose from airborne noble gases and
fission (plus activation) products; (2) doses from radionuclides deposited on
the ground or in water; and (3) dose due to direct exposure to radioactive
material at the facility, including nitrogen-16 in turbine buildings (in boiling
water reactor plants) and other radionuclides in stored wastes.
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373
APPENDIX I
I.1.1 Dose from Airborne Noble Gases and Fission Products
Estimates of nuclide-specific ground-level activity concentrations in air
at a particular direction and distance and annual and quarterly doses can be
calculated as a function of distance and direction using accepted air disper-
sion methodologies that account for radioactive decay and plume depletion
during transport, release height, and average annual (or longer) meteorol-
ogy (wind speed, direction, atmospheric stability) as well as site-specific
features such as terrain features. The organ dose resulting from immersion
in air containing radioactive gases (sometimes referred to as a radioactive
plume) at any location can then be calculated fairly accurately for each of
the specific nuclides released and their specific gamma and beta emissions
(Federal Guidance Report 12 [USEPA, 1993]).
The exposure rate from immersion in a plume of noble gases varies
significantly with the composition of the gaseous cloud versus distance. The
exposure rate from the various radioactive gases varies significantly due to
large differences in the energies of their respective radiation emissions. As
shown in Table I.1, the effective dose factors for short-lived emitters such
as krypton-87 and 88 are significantly higher than that for longer-lived
xenon-133, which comprises most of the airborne effluents from currently
operating nuclear plants. However, because of the shorter half-lives of these
radioisotopes, their relative contribution to doses to persons living farther
downwind will be somewhat less than the relative effective dose factors
shown in Table I.1.
I.1.2 Doses from Deposited Radionuclides
Calculations of external exposure and organ doses from particulate
radioactive materials deposited on the ground are based on the same trans-
port model used for estimating noble gas concentrations downwind and
models for calculating dry and wet deposition and the dose rate per unit
TABLE I.1 Exposure Rate Dose Conversion
Factors
Effective Dose Factor
(Sv Bq–1 s m–3)
Nuclide Half-life
10–14
Kr-87 76 min 4.0 ×
10–14
Kr-88 2.8 h 9.7 ×
10–15
Xe-133 5.2 d 1.3 ×
10–14
Xe-135 9.1 h 1.1 ×
10–14
Xe-135m 15 min 1.9 ×
10–14
Xe-138 14 min 5.5 ×
SOURCE: Effective dose factors from Federal Guidance
Report 12.
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374 APPENDIX I
activity concentration in soil of each nuclide. Recommended models for
estimating doses from external radiation exposure are discussed in USNRC
regulatory guides as well in guidance published by the U.S. Environmental
Protection Agency (USEPA) and the National Council on Radiation Pro-
tection and Measurements (NCRP). Some of these models conservatively
assume that the activity is deposited onto the surface of the ground (no
ground roughness correction) and that no weathering occurs to reduce the
integral exposure over time. Nevertheless, the estimated doses from nuclear
plant effluents are a small fraction of those resulting from immersion in the
plume of noble gases that are released from the plants, and they are almost
always too low to be measured directly.
The exposure rate from radionuclides deposited onto the ground varies
with the energy of the emissions. However, longer-lived nuclides can build
up in the soil with time. Table I.2 shows conservative estimates of exposure
in air per unit surface activity concentrations for selected radionuclides of
importance in airborne effluents. The tabulated values are for a plane sur-
face source. The exposure rates for a given activity concentration in the soil
will decrease as the activity moves down into the soil profile over time as a
result of rainfall and human activity. Because of the very low effluent rates
and the diffusion of the airborne activity over a large area, only the longer-
lived nuclides such as cesium-137 and cobolt-60 can potentially build up
to activity levels high enough for the exposure rate to be distinguishable
from even the temporal variations in terrestrial background levels at any
site. Modern gamma-ray spectrometric techniques might allow the detec-
tion of very low levels of cobalt-60 in soil at close-in sites that might occur
after many years of plant operation, but cesium-137 from the facility, even
if present, would be undetectable because it is expected to be present in all
soils from nuclear weapons testing fallout.
TABLE I.2 Exposure Rate per Unit
Deposition Density
Exposure Rate
(µR/h per nCi/m2)
Nuclide Half-life
I-131 8d 0.0073
Cr-51 28 d 0.0006
Co-60 5y 0.0432
Cs-134 2.1 y 0.0291
Cs-137 30 y 0.0107
Ba-140 12 d 0.0027
SOURCE: Beck (1980).
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375
APPENDIX I
I.2 INTERNAL EXPOSURES
Calculation of radiation doses from internally deposited radionuclides
is done by determining the spatial and temporal distribution of energy
deposited in tissues and organs after intake. Generally, this requires knowl-
edge of the distribution of sources and targets in space and time. The source
is the radionuclide of concern, and the target is the biological entity consid-
ered most relevant for determining dose and risk. The choice of target can
range from molecules and cells to organs and tissues to whole organisms.
For radiation protection, the level of averaging of radiation dose has con-
sistently been at the tissue or organ level.
Regardless of dosimetry system employed, the following information
is needed:
• decay characteristics of the radionuclide,
• chemical and physical nature of the exposure material,
• intake route,
• solubility of the exposure material in vivo,
• tissue and organ distribution pattern in the body,
• retention times for the radionuclide in the various target tissues,
and
• an appropriate anatomic or physiologic model of a human.
Taken together, this information allows both dose rate and dose pat-
terns from intakes of radionuclides to be calculated.
For calculating internal doses resulting from the release of radionu-
clides from nuclear facilities, the USNRC continues to use dosimetry meth-
ods published by the International Commission on Radiological Protection
in 1959 (commonly referred to as ICRP 2 methods) (ICRP, 1960). This
is described in USNRC Regulatory Guide 1.109 (USNRC, 1977), which
implements the guidance in Appendix I of 10 CFR Part 50. The ICRP
2 dosimetry model (ICRP, 1960) was developed primarily for providing
radiation protection guidance for occupational environments, although
recommendations for members of the public living in the neighborhood of
controlled areas are also provided. However, the ICRP recommendations
for the public did not take into account differences in dose limits between
workers and members of the public, nor did they use different biokinetic
models; thus, the differences in maximum permissible concentrations only
reflect different exposure periods, that is, 40-hour weeks for workers versus
168-hour weeks for the public.
In general, the guidance protects workers by controlling the dose to the
“critical organ,” which is defined as that organ of the body that receives
the highest dose or is the most radiosensitive organ receiving a significant
dose from an intake of a given radionuclide.” Through the use of the critical
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376 APPENDIX I
organ and maximum permissible doses defined in ICRP 2, the risk to the
individual is then controlled through the use of the Maximum Permissible
Body Burden (q). This quantity is applied to specific exposure scenarios
(e.g., chronic exposure for 168 hours per week for 50 years) and used with
defined anatomic and physiologic parameters for ingestion and inhalation
to yield Maximum Permissible Concentrations for a radionuclide in air
(MPCa) and water (MPCw). Although the USNRC does not use the dose
constraints proposed in ICRP 2, but rather those in 10 CFR 50 Appendix I,
it still uses the ICRP 2 methodology for calculating doses to the maximally
exposed member of the public.
The models used in ICRP 2 to define intakes from ingestion and in-
halation exposure to radionuclides are very basic, reflecting the state of
knowledge of the behavior of radionuclides at the time this methodology
was issued. All physiologic parameters were provided for a Standard Man,
and thus do not provide for individual variations in body size, intake, or
metabolic rates. The Standard Man, which was defined at the Tripartite
Conference in Chalk River (Warren et al.,1949), was designed to represent
a typical or average adult who is exposed occupationally. Although the
USNRC has modified the application of the Standard Man approach as
applied to intake of radionuclides in effluents from nuclear plants, the es-
sential features of Standard Man are described here for reference.
Water balance in Standard Man is defined in terms of food, fluids,
and oxidation by-products intake and excretion rates, as shown Table I.3.
Other physiologic parameters were also defined (Table I.4). These values
allow the calculation of intakes from ingestion and inhalation in terms of
the quantity of radionuclides in food, water, and air. In addition, a sepa-
rate empirical model was defined for intakes of particulates by inhalation
(Table I.5). Although it was recognized that the retention of particulate
matter depends on many factors, such as the size, shape, and density of the
particles, as well as their chemical form and whether the person is a nose
or mouth breather, ICRP indicated that specific data were lacking, and
therefore the distribution and fate of inhaled particles could adequately be
described as in Table I.5.
Thus, there is no particle size dependence in this model, which strongly
affects both total and regional deposition in the respiratory tract. Addition-
ally, the fate of material, whether being cleared via feces as particles or
absorbed to blood, was described simply in terms of whether the inhaled
particles were relatively soluble or not. For the soluble compounds, the
25 percent deposited in lungs was assumed to translocate to blood within
the first 24 hours after inhalation. For the insoluble particles, half of the
25 percent that deposited in the lung was assumed to be eliminated from
lung and swallowed in the first 24 hours after inhalation; this meant that
62.5 percent of the materials deposited in the upper respiratory tract (URT)
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APPENDIX I
TABLE I.3 Intake and Excretion of Standard Man
(Water Balance)
Intake (cm3/d) Excretion (cm3/d)
Food 1000 Urine 1400
Fluids 1200 Sweat 600
Oxidation 300 From lungs 300
Feces 200
TOTAL 2500 2500
SOURCE: Warren et al. (1949).
TABLE I.4 Other Physiologic Parameters for Standard Man
Vital capacity of lung (male) 3-4 L
Vital capacity of lung (female) 2-3 L
2 × 107 cm3 d–1
Air inhaled per 24-h day
50 m2
Interchange area of lungs
20 m2
Area of upper respiratory tract, trachea and bronchi
70 m2
Total surface area of respiratory tract
4.3 × 104 g
Total water in body
Average lifespan of man 70 y
SOURCE: Warren et al. (1949).
TABLE I.5 Behavior of Inhaled Particulates in the Respiratory Tract of
Standard Man
Distribution Readily Soluble Compounds Other Compounds (insoluble)
Exhaled 25 25
Deposited in URT and 50 50
swallowed
Deposited in lungs (LRT) 25 (to blood) 25 (12.5% swallowed; 12.5%
to blood)
SOURCE: Warren et al. (1949).
and lower respiratory tract (LRT) was removed by mucociliary clearance,
swallowed, and subsequently would be excreted via feces. The remaining
12.5 percent of the amount deposited in LRT was absorbed to blood with
a 120-day half-time.
To calculate the absorbed doses, the retention and fate of a radionuclide
taken into the body by ingestion or inhalation had to be described for in-
dividual radionuclides once they reached the blood. To do this for most of
the radionuclides, particularly those for which the bone and GI tract were
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378 APPENDIX I
not the critical organs, a simple exponential model was assumed as default.
This was expressed by the equation:
qf2 = P(1 – e–λt)/λ (1)
where
qf2 = amount of the radionuclide in the critical body organ (Ci)
f2 = fraction of radionuclide in the critical organ to that in total
body
λ = effective decay constant = 0.693/T
T = effective half-time ((TrTb)/(Tr + Tb) (days)
Tr = radiological half-time (days)
Tb = biological half-time (days)
T = period of exposure (for occupational exposure, t = 50 years)
P = rate of uptake of the radionuclide by the critical body organ
(µCi d–1)
The quantities and their radionuclide-specific values needed to calculate
the absorbed and rem doses were provided in Table 12 of ICRP 2 (1960)
and included the reference organ for dose calculation; the physical, bio-
logical, and effective half-times; the fraction of ingested radionuclide that
reached the blood (f1); the critical organ fraction (f2); and the fractions
reaching the critical or reference organ from water (fw) or air (fa).
ICRP dosimetry models have been improved markedly since the release
of ICRP 2, and the models used in ICRP 2 have been replaced by more
current dosimetry models. These models have been designed to calculate
age-dependent dose coefficients (dose per unit intake) for members of
the public. These include doses from ingestion (ICRP, 1989, 1993) and
inhalation (ICRP, 1995a,b), doses to the embryo and fetus from radionu-
clide intakes by the mother (ICRP, 2001), doses to infants from ingestion
of mothers’ milk (ICRP, 2004), a new respiratory tract dosimetry model
(ICRP, 1994), and an alimentary tract dosimetry model (ICRP, 2006).
Also contained in the above documents are radioelement-specific biokinetic
models that describe the systemic tissue and organ uptake and retention
of radionuclides once they have reached the blood. These systemic models
are coupled with the appropriate intake model (ingestion, inhalation) and
a dosimetric model that calculates the dose to all target organs and tissues
per radionuclide decay to obtain exposure-specific dose coefficients.
I.2.1 ICRP Models to Support an Epidemiologic Study
The first internal dosimetry system was published in 1959 (ICRP, 1960)
and has generally been replaced sequentially by the ICRP 30-based sys-
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379
APPENDIX I
tem (ICRP, 1979), which was focused on occupational workers; the ICRP
56-based system (ICRP, 1989), which related to members of the public; and
the current system outlined (but not described) in ICRP 103 (2007). Al-
though the ICRP 2 system is still implemented by USNRC for performance
compliance dosimetry of radioactive effluent releases from nuclear plants
in some ways, it has been described previously. Rather, the more current
ICRP (ICRP 56+) dosimetry system (ICRP, 1990, 1992, 1995a,b), which
may be most applicable for calculating doses for an epidemiologic study,
is described below.
Over the past 50 years, a substantial increase in knowledge about ra-
dionuclide metabolism and biokinetics in humans and experimental animal
models has occurred and has provided a basis for the development of more
realistic biokinetic models of radionuclide uptake and retention, particu-
larly at the organ and tissue level. This plus better understanding of the
disposition of inhaled and ingested radionuclides both in the deposition and
systemic organs has further provided the basis for significant improvements
in internal dosimetry and modeling.
The current generation of ICRP models for internal dosimetry of in-
takes of radionuclides by the public offers the following advantages:
1. More complete radionuclide physical decay schemes;
2. Improved physical anthropometric models, which allow more ac-
curate calculation of absorbed fractions of radiation resulting from
the distribution of radionuclides in various source organs;
3. Better description of organ-level biokinetics of radionuclides that
reach the blood and circulation (systemic models);
4. More anatomically and physiologically accurate model of the re-
spiratory tract together with improved description of deposition,
retention, translocation, and excretion of inhaled radionuclides;
5. More anatomically and physiologically accurate model of the ali-
mentary tract, which extends the number of tissues modeled and
includes a better understanding of the biokinetics within the ali-
mentary tract and relative radiosensitivities of the various target
tissues within the alimentary tract;
6. Improved age-dependent modeling of radionuclide biokinetics in
humans of different ages;
7. Addition of radionuclide biokinetic modeling of the uptake and
retention of radionuclides in the embryo and fetus from intakes by
the mother, both before and during pregnancy;
8. Inclusion of a milk pathway of intake for newborns who are nursed
by mothers who have had intakes of radionuclides.
These improvements in modeling have necessarily come at the expense
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380 APPENDIX I
FIGURE I.1 Example compartmental model representation of a radionuclide bio-
kinetic model SOURCE: Adapted igure I.1.eps
F from ICRP (1997).
bitmap
of having much more complicated models, which require the use of com-
puter software to calculate radiation doses.
Figure I.1 shows an example of the type of biokinetic models being
used by ICRP in its current series of dosimetry models. Among the general
features of the modeling approach are (1) allowance of intakes by inges-
tion, inhalation, wounds, and transcutaneous absorption across intact skin.
(2) Compartments in brown are tissue sites of entry of radionuclide into
the body. These may be described in more detail in other models, e.g., a
respiratory tract model. (3) Compartments in blue are systemic deposition
sites that communicate directly with blood. (4) Current models allow for
recycling between compartment, which can be a more accurate repre-
sentation of the flow of radionuclides between compartments. Different
levels of subcompartments within a tissue compartment can also be used
when multicomponent retention patterns are needed. For example, multiple
compartments have been employed for the liver in the plutonium systemic
biokinetic model of ICRP publication 67 (1992).
The complexity of a given set of biokinetic models depends on the
tissues and organs that are the principal deposition sites for a given ra-
dionuclide, and are therefore usually at greater risk of receiving radiation
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381
APPENDIX I
dose. When designing the models, a full range of radionuclides is consid-
ered. Additionally, the list of tissues and organs is also influenced by those
considered to be at risk of biological effects from radiation. Since this list
includes irradiation from both external and internal sources, essentially all
tissues and organs of the body are considered. ICRP Publication 103 (2007)
lists the following organs: red bone marrow, colon, lung, stomach, breast,
gonads, urinary bladder, esophagus, liver, thyroid, bone surface, brain,
salivary glands, skin, adrenals, extrathoracic region of the respiratory tract
(head airways), gall bladder, heart, kidneys, lymph nodes, muscle, oral mu-
cosa, pancreas, prostate, small intestine, spleen, thymus, and uterus/cervix.
To calculate organ-specific absorbed doses, two biokinetic models are
required. The first model is used to relate radionuclide concentration in air
or solid media (food or water) to intake. This is done using the Human
Respiratory Tract Model (HRTM) (ICRP, 1994) or the Human Alimentary
Tract Model (HATM) (ICRP, 2006) for inhalation and ingestion, respec-
tively. The second model, which is radioelement specific, is the systemic
biokinetic model, which describes in detail the spatial and temporal distri-
bution of a radionuclide once it has reached the blood. These models are
coupled mathematically so that the number of disintegrations occurring in
the various organs and tissues of interest can be calculated and used to-
gether with an appropriate anatomical model and physical dosimetry model
to calculate the pattern of deposition of energy in the organs.
I.2.1.1 Human Respiratory Tract Model
The HRTM is actually a second-generation replacement of the simple
respiratory tract model published in ICRP 2 (1960); it replaced the inter-
mediate model published in ICRP Publication 30 (1979). The HRTM was
developed by ICRP over a 10-year period and represented the state-of-
the-art knowledge about the behavior of inhaled particles and gases in the
human respiratory tract. In this model, the respiratory tract is subdivided
into five anatomical compartments (Figure I.2), ranging from two extra-
thoracic regions (ET1, ET2), to bronchi, bronchioles, and the parenchymal
region of the lung (AI). Regional deposition efficiencies were calculated
for these anatomic compartments for particle sizes ranging from 0.001 µm
through 100 µm. As part of the definition of these anatomic regions, dif-
ferent geometric constructs were created for each of the regions. The criti-
cal cell populations at risk to stochastic health effects were purported to
exist within these geometrically prescribed subregions so that only energy
deposited in these subregions is used to calculate the absorbed dose to that
anatomic compartment. Additionally, each of the anatomic compartments
has been risk-weighted by apportioning the radiation detriment to the dif-
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382 APPENDIX I
FIGURE I.2 HRTM anatomic model. SOURCE: ICRP (1994).
Figure I.2.eps
bitmap
ferent compartments based on human and experimental data regarding the
frequency of different types of respiratory cancer (ICRP, 1994).
The fate of deposited radionuclides in the respiratory tract1 is modeled
by considering clearance based on three pathways: (1) mucociliary clear-
ance from both the head airways and the lung leading to swallowing of the
cleared material, and subsequent excretion into feces or absorption through
the GI tract to blood; (2) clearance of particles through the interstitium
leading to uptake in the lymph nodes that drain the various regions of the
respiratory tract; and (3) dissolution of the radionuclide on or near the
airways of the respiratory tract followed by either local binding to tissue
constituents (less likely and applicable to only a few radioelements, e.g.,
plutonium and americium) or absorption to blood (most likely). These pro-
cesses are modeled mathematically as competing pathways and are depen-
dent on the physical and chemical properties of the inhaled radionuclide.
The HRTM is an age-dependent dosimetry model whose morphometric
and physiologic characteristics have been defined for reference ages of 3
months; 1, 5, 10, and 15 years; adult; and all for both genders. As such,
age-dependent dose coefficients (dose per unit intake) have been published
for members of the public in ICRP Publication 71 (1995). The model also
1 It is important to note that not all inhaled material is deposited in the respiratory tract.
About 40-50 percent of most inhaled material is exhaled without depositing anywhere.
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383
APPENDIX I
has examined the role of personal factors such as smoking and respiratory
disease on deposition and clearance of inhaled particles, both of which af-
fect the dose coefficients.
Because of the complexity of the HRTM, several software programs
have been developed and implement the model for use in dose assessment
and bioassay interpretation (e.g., Bertelli et al., 2008; Jarvis and Birchall,
1994).
I.2.1.2 Human Alimentary Tract Model
The HATM (ICRP, 2006) is a biokinetic and dosimetric model of the
human alimentary tract that replaces the previous GI tract model of ICRP
Publication 30 (1979). This expanded model is applicable to all radionu-
clide intakes by children and adults. As such, it provides age-dependent
parameter values for the dimensions of the alimentary tract as well as age-
and gender-dependent transit rates. Although the default is for absorption
of radionuclide to blood to occur in the small intestine, the model does
allow for absorption to occur in other regions. The HATM also allows
for local binding of radionuclides to the structures of the various regions
of the alimentary tract, thus allowing for calculation of radiation dose to
subcompartments of the HATM.
Figure I.3 illustrates the compartmental model for the HATM. It de-
picts the entire alimentary tract from oral cavity to rectosigmoid colon.
Input occurs into the oral cavity via ingestion and clearance of inhaled
deposited radionuclides from the respiratory tract into the esophagus (the
HATM was designed to be consistent with the HRTM in terms of structure,
clearance processes, and dosimetric modeling). The movement of contents
through the alimentary tract is sequential, and the transit rates are mod-
eled by first-order exponential processes. It was recognized that modeling
transit in this way was a considerable simplification, but by indexing the
emptying half-time to the reported mean transit times of material through
a given segment, a reasonable estimate of the transit rate was obtained,
which allowed dose calculation to be done in a straightforward way. The
bulk of the material ends up in feces. It should be noted that the behavior
of a given radionuclide in terms of absorption to blood versus excretion in
the bulk material depends on its physical and chemical form.
Absorption of solutes, including radionuclides, to blood is allowed
through the walls of all HAT organs, but the default is that absorption is
limited through the small intestine. Deposition and retention of radionu-
clides in teeth, oral mucosa, and GI tract walls allows these tissues to be
both sources for radionuclide retention and targets for calculating radiation
absorbed doses, although these tissues are targets in any case. Typically, the
geometry identified for dose calculations in the various tissues of the HAT
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384 APPENDIX I
FIGURE I.3 Compartments of the HATM. SOURCE: ICRP (2006).
Figure I.3.eps
bitmap
comprise the layers of epithelial cells contained in those tissues. This is due
to the fact that most of the cancers linked to radiation in the alimentary
tract are epithelial in origin.
Transit time parameter values have been provided for different types
of ingested materials (solids, caloric and noncaloric liquids, and total diet)
and for subjects having ages of 3 months, 1 year, 5-15 year, adult male,
and adult female.
I.2.1.3 Systemic Biokinetic Models
The development of radioelement-specific systemic biokinetic models
is ongoing within the committees and task groups of the ICRP. Presently
the only relatively complete set of systemic models, i.e., for all radioele-
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APPENDIX I
ments, is that contained in the ICRP 30 series of publications, which apply
only to adult workers. From the structural point of view, these models are
nonrecycling models whose physiologic relevance is often questionable, but
they are useful for their intended purpose of designing radiation protection
programs as well as interpretation of human bioassay data.
A smaller number of age-dependent, recycling biokinetic models have
been published by ICRP, namely for the alkaline earths and lead, calcium,
plutonium, neptunium, americium, and curium. Age-specific biokinetic
data have been developed for the most common radionuclides (isotopes of
hydrogen, carbon, zirconium, niobium, ruthenium, iodine, cesium, cerium,
plutonium, americium, and neptunium) for a total of 31 radioelements
(ICRP, 1989). Recycling models continue to be developed by ICRP for other
radioelements, but these may not become available during the timeframe
needed for this project. Nevertheless, ICRP in its publication 72 (1995a)
added 60 other radioelements to its age-specific dose coefficient database
by using the nonrecycling models of ICRP publication 30 together with
age-specific organ masses.
ICRP publication 72 (ICRP, 1995a) provides age-specific dose coef-
ficients that are needed for the purposes of epidemiologic study dosimetry.
Although ICRP states “[b]ecause changes in biokinetics are considered with
age and have not been considered fully, these additional dose coefficients
should be used with care for assessing doses to infants and children,” the
dose coefficients nevertheless provide the best set of age-dependent dose
coefficients available. Additionally equivalent doses have been provided in
electronic form by ICRP on CD.
I.2.1.4 Comparison of USNRC and Recent ICRP Dose Coefficients
In Table I.6, the inhalation dose coefficients from USNRC Regulatory
Guides are compared with those derived from recent ICRP publications
(ICRP, 1995b) for radionuclides commonly encountered in effluent releases
from nuclear power plants. It is clear that very large differences are ob-
served between the two sets of data. USNRC dose coefficients are derived
from USNRC Regulatory Guide 1.109 (1977), Table E-7. The tabulated
values were converted to Sv/Bq from mrem/pCi by dividing the latter by
3700.
ICRP dose coefficients were calculated using the AIDE dose assessment
software (Bertelli et al., 2008). All coefficients were calculated assuming an
aerosol particle size of 1.5 µm AMAD, inhaled by a male worker. Solubility
classes (F or M) are shown in the radionuclide column. The systemic models
were derived either from ICRP 56 or ICRP 67.
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386 APPENDIX I
TABLE I.6 Comparison of Inhalation Dose Coefficients (Committed
Dose per Unit Uptake) Derived from USNRC and ICRP Dosimetric
Approaches for Adults (Sv/Bq intake)
GI-LLIa
Radionuclide Model Bone Liver Lung
H-3 USNRC — 4.27E-11 4.27E-11 4.27E-11
H-3 (F) ICRP (56) 8.14E-12 8.14E-12 8.17E-12 8.61E-12
Co-60 USNRC — 3.89E-10 2.02E-7 9.62E-9
Co-60 (M) ICRP (67) 3.72E-9 8.11E-9 4.89E-8 5.56E-9
Sr-90 USNRC 3.35E-6 — 3.24E-7 2.43E-8
Sr-90 (F) ICRP (56) 3.63E-7 6.58E-10 7.12E-10 1.21E-8
Ru-106 USNRC 2.28E-9 — 3.16E-7 3.08E-8
Ru-106 (M) ICRP (30) 2.71E-9 2.82E-9 1.76E-7 2.55E-8
I-131 USNRC 8.54E-10 1.21E-9 4.03E-7 2.12E-10
(Thyroid)
I-131 (F) ICRP (56) 5.95E-11 2.01E-11 1.76E-7 4.04E-11
(Thyroid)
Cs-137 USNRC 1.61E-8 2.10E-8 2.54E-9 2.85E-10
Cs-137 (F) ICRP (56) 5.60E-9 5.52E-9 5.19E-9 6.76E-9
aGI-LLI, gastrointestinal tract–lower large intestine.
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