I

Radiation Dose Assessment

Under normal operating conditions, nuclear facilities release radioactive 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 decay). 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 ingestion 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.


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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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377 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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385 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. REFERENCES Beck H.L. (1980), Exposure rate conversion factors for radionuclides deposited on the ground, Environmental Measurements Laboratory report EML-378, U.S. Department of Energy, Environmental Measurements Laboratory, New York, NY. Bertelli, L, D. R. Melo, J. Lipsztein, and R. Cruz-Suarez (2008). AIDE: Internal dosimetry software. Radiat. Protect. Dosim. 130:358-367. ICRP (International Commission on Radiological Protection) (1959). Permissible Dose for Internal Radiation, ICRP Publication 2 1959 Superseded by ICRP Publication 30. ICRP (1960). Report of ICRP Committee II on Permissible Dose for Internal Radiation (1959), with Bibliography for Biological, Mathematical and Physical Data. Health Phys. 3:1-380. ICRP (1979). Limits for Intakes of Radionuclides by Workers. ICRP Publication 30. Ann. ICRP 2(3-4). ICRP (1989). Individual Monitoring for Intakes of Radionuclides by Workers, ICRP Publica- tion 54. ICRP (1990). Age-dependent Doses to Members of the Public from Intake of Radionuclides— Part 1. ICRP Publication 56. Ann. ICRP 20(2). ICRP (1992). Age-dependent Doses to Members of the Public from Intake of Radionuclides— Part 2 Ingestion Dose Coefficients. ICRP Publication 67. Ann. ICRP 22(3-4). ICRP (1993). Protection Against Radon-222 at Home and at Work. ICRP Publication 65. ICRP (1994). Human Respiratory Tract Model for Radiological Protection. ICRP Publication 66. Ann. ICRP 24(1-3). ICRP (1995a). Age-dependent Doses to the Members of the Public from Intake of Radio- nuclides—Part 5 Compilation of Ingestion and Inhalation Coefficients, Publication 72. ICRP (1995b). Age-dependent Doses to Members of the Public from Intake of Radionuclides— Part 4 Inhalation Dose Coefficients. ICRP Publication 71. Ann. ICRP 25(3-4). ICRP (1997). Individual Monitoring for Internal Exposure of Workers. Replacement of ICRP Publication 54. ICRP Publication 78. Ann. ICRP 27(3-4).

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387 APPENDIX I ICRP (2001). Doses to the Embryo and Fetus from Intakes of Radionuclides by the Mother. ICRP Publication 88. ICRP (2004). Doses to Infants from Ingestion of Radionuclides in Mothers’ Milk. ICRP Pub- lication 95. Ann. ICRP 34(3-4). ICRP (2006). Human Alimentary Tract Model for Radiological Protection. ICRP Publication 100. Ann. ICRP 36(1-2). ICRP (2007). The 2007 Recommendations of the International Commission on Radiological Protection. ICRP Publication 103. Jarvis, N., and A. Birchall (1994). LUDEP 1.0, a personal computer program to implement the new ICRP respiratory tract model. Radiat. Protect. Dosim. 53:191-194. USEPA (U.S. Environmental Protection Agency) (1993). External Exposure to Radionuclides in Air, Water, and Soil. Federal Guidance Report No. 12 EPA-402-R-93-081. Oak Ridge National Laboratory, Oak Ridge, TN Washington, DC: USEPA. USNRC (U.S. Nuclear Regulatory Commission) (1977). Regulatory Guide 1.109. Calculation of Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluat- ing Compliance with 10 CFR Part 50, Appendix I. Revision 1. October. Warren, S., A. C. Chamberlain, G. J. Neary, E. F. Edson, G. O. Failla, J. C. Hamilton, L. Hempelman, H. M. Parker, K. Z. Morgan, B. S. Wolf, A. Brues, L. S. Taylor, W. Langham, D. Hoffman, W. B. Lewis, A. J. Cipriani, G. C. Laurence, H. Carmichael, G. H. Guest, E. Renton, G. E. McMurtrie, and A. O. Bratten (1949). Minutes of the Permissible Doses Conference Held at Chalk River, Canada, September 29-30, R.M.-10, Tri-Partite Conference, Chalk River, Canada.

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