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3 Radiation Dose Assessment T his chapter addresses the first charge in the statement of task for this study (see Sidebar 1.1 in Chapter 1) on methodological approaches for assessing offsite radiation doses to populations near nuclear plants and fuel-cycle facilities in the United States. It is specifically intended to address the following issues: • Pathways, receptors, and source terms. • Approaches for overcoming methodological limitations arising from the variability in radioactive releases over time as well as other confounding factors. • Approaches for characterizing and communicating uncertainties. Information on the availability, completeness, and quality of radioactive ef- fluent releases from nuclear facilities, which is also part of this first charge, was addressed in Chapter 2. 3.1 BACKGROUND ON DOSE ASSESSMENT AND DOSE RECONSTRUCTION When ionizing radiation interacts with the human body it transfers part or all of its energy to the molecules and cells of body tissues. The response of these tissues to the deposition of energy in terms of physical, chemical, and biological changes is dependent on the amount of energy deposited per unit mass of tissue, or absorbed dose (see Table 3.1). The quantity absorbed dose (D) is defined as the mean energy imparted by ionizing radiation per 97
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98 ANALYSIS OF CANCER RISKS TABLE 3.1 Selected Quantities and Units for Radiation Exposure and Dose SI Unit or Its Special Relationship Field of Quantity Old Unit Name Between Units Application Reference C kg–1 1 R = 2.58 10–4 Exposure R Monitoring NCRP C kg–1 (2007) Absorbed dose rad Gy 1 rad = 0.01 Gy Research ICRP (2007b) Dose equivalenta rem Sv 1 rem = 0.01 Sv Radiation ICRP Protection (1977) Equivalent dosea rem Sv 1 rem = 0.01 Sv Radiation ICRP protection (1991) Effective dose rem Sv 1 rem = 0.01 Sv Radiation ICRP equivalentb protection (1991) Effective doseb rem Sv 1 rem = 0.01 Sv Radiation ICRP protection (1991) Committed effective rem Sv 1 rem = 0.01 Sv Radiation ICRP dose equivalentc protection (1991) (CEDE) Collective dose person-rem person-Sv 1 person-rem = Radiation ICRP equivalent 0.01 person-Sv protection (1991) aDose equivalent and equivalent dose are conceptually similar. However, dose equivalent makes use of quality factors (QFs), which were replaced with radiation-weighting factors (w R) for the calculation of equivalent doses. bEffective dose equivalent and effective dose are conceptually similar. Effective dose equiva- lent is the weighted sum of the dose equivalents over all organs and tissues of the body, using tissue-weighting factors (wT), whereas effective dose is the weighted sum of the equivalent doses over all organs and tissues of the body. An additional difference is that different wT values are used in the calculation of effective dose equivalent and effective dose. cCommitted effective dose equivalent is the time integral of the effective dose equivalent from the time of the activity intake until the age of 70 y. unit mass at a point of interest. The unit of absorbed dose is J/kg, and its special name is the gray (Gy) (ICRU, 2011). Although defined as a point quantity, absorbed dose usually represents an average over some finite volume or mass, such as the mass of the thyroid or the volume of red bone marrow distributed in the entire body. When the absorbed dose has ap- proximately the same value for all organs and tissues of the body, as is the
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99 RADIATION DOSE ASSESSMENT case for direct radiation1 from energetic gamma rays or internal irradiation from inhalation or ingestion of cesium-137, it is common to use the term whole-body absorbed dose. The quantity referred to as dose equivalent (HT) is also used in some dose calculations, for example, for calculating doses to the maximally ex- posed individual, or MEI2 (USNRC, 1977a) around nuclear facilities (see Table 3.1). Dose equivalent is defined as the absorbed dose modified by a quality factor (QF) that represents the relative biological effectiveness of a radiation type: HT = D × QF (1) In the U.S. Nuclear Regulatory Commission’s (USNRC’s) fundamental regulatory radiation protection guidance (10 CFR Part 20, Standards for Protection Against Radiation), QF takes on values of unity (1) for X rays, gamma rays, and beta radiation; 20 for alpha particles, fission fragments, and heavy particles of unknown charge; and 10 for high-energy protons and neutrons of unknown energy. More recent radiation protection guidance from the International Com- mission on Radiological Protection (ICRP) defines other dose quantities. These include equivalent dose and effective dose (ICRP, 1991; see Table 3.1). As radiation protection guidance has evolved over the years, the appli- cation of various dose quantities has become more clearly prescribed. For example, as stated in ICRP Publication 103 (2007b): The main and primary uses of effective dose in radiological protection for both occupational workers and the general public are: • rospective dose assessment for planning and optimization of protec- p tion; and • etrospective dose assessment for demonstrating compliance with dose r limits, or for comparing with dose constraints or reference levels. Thus, effective dose and equivalent dose have been used for regulatory 1 As noted in Chapter 2, direct radiation exposure refers to external whole-body radiation exposure from ionizing radiation emitted by radionuclides in the air, soil, sediments, or water bodies as well as radiation from sources within the site boundary. The latter includes radioac- tive wastes buried or stored onsite as well as N-16 produced in the turbines of boiling water reactors. 2 MEI is a regulatory construct for assessing compliance with radiation protection standards. It refers to a hypothetical individual who is postulated to receive the maximum possible radiation dose from a facility because of his or her location relative to the facility as well as lifestyle habits.
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100 ANALYSIS OF CANCER RISKS purposes worldwide, and the latter is used in the current USNRC dose com- pliance formalism. In essence, the calculation of effective dose for external exposure, as well as dose coefficients for internal exposure, are based on absorbed dose, weighting factors, and reference values for the human body and its organs and tissues. In general, effective and equivalent doses do not provide individual-specific doses, but rather doses for a reference person3 (such as an MEI) under a given exposure situation. Effective and equivalent doses, as well as collective dose4 (see Table 3.1), were not designed for research purposes. Consequently, the use of these quantities should be avoided in epidemiologic studies because they mask many uncertainties that are embedded in their formalism, for ex- ample, uncertainties in radiation and tissue weighting. It is prudent to use the more fundamental dose quantity, D, for dose assessments used in epidemiologic studies. For such studies, absorbed dose is usually estimated for specific organs on an annual basis, expressed as rad/yr. In the context of this discussion, the term dose assessment refers to the estimation of absorbed doses received by individuals as a result of exposure to ionizing radiation. Absorbed doses from direct radiation ex- posure5 can be estimated using equipment that measures exposures in air in real time, for example by using radiation-sensitive materials such as thermoluminescent detectors (TLDs). Alternatively, doses can be estimated retrospectively by reconstructing an individual’s past exposure to ionizing radiation. Absorbed dose from internal exposure (i.e., inhalation, ingestion, or absorption of radionuclides) can be estimated from measurements of radionuclide concentrations in air, soil, and food. Both exposure and dose can be estimated using models that relate releases of radioactivity to the environment (e.g., facility effluents) to exposure rates in air and to radio- nuclide concentrations in air, water, and food. Dose reconstruction is the primary concern of this chapter. Reconstructing an individual’s absorbed dose from releases of radioac- tive effluents from a nuclear plant or fuel-cycle facility requires knowledge of several factors, including: 3 The most recent ICRP guidance (ICRP 101) uses the term “representative person” in- stead of “reference person” (ICRP, 2007a). However, the USNRC continues to use the older terminology. 4 Collective dose is the sum of individual doses received by a specified population over a specified period of time. Collective dose is sometimes referred to as the population dose. ICRP (2007b) notes that collective dose is a useful concept for radiological protection but is not appropriate for use in epidemiologic studies or risk projections. 5 As shown in Table 3.1, radiation exposures are expressed in terms of Roentgen (R). In the 1970s, it was common practice to convert exposure measurements in R to absorbed doses in air in rad using the conversion factor 1 R = 0.875 rad.
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101 RADIATION DOSE ASSESSMENT • Amount of radioactive material released from a facility, or source term; • Transport of this radioactivity through the environment; and • Uptake of (or exposure to) this radioactivity by the individual. There are many pathways by which individuals can be exposed to ra- diation, be it from naturally occurring or manmade sources. As illustrated in Figure 3.1, individuals can be exposed to: • External radiation from radionuclides that emit penetrating ra- diation (i.e., high-energy radiation such as gamma radiation that penetrates the human body). This radiation can be received directly FIGURE 3.1 Pathways for exposure to radiation from effluent releases from nuclear plants and fuel-cycle facilities. SOURCE: 3.1.eps al. (1974). Figure Soldat et bitmap
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102 ANALYSIS OF CANCER RISKS from a facility, from radionuclides present in air, or from radionu- clides deposited on the ground or in local water bodies. External exposure is usually the principal exposure route for radioactive effluent releases from nuclear plants. • Internal radiation from radionuclides that are inhaled, ingested, or absorbed through intact or broken skin. Ingestion is usually the principal route of intake for radioactive effluent releases associated with nuclear fuel-cycle facilities. Sophisticated computer models have been developed to reconstruct doses from exposures to external and internal radiation. To estimate exter- nal dose, transport calculations are carried out to determine atmospheric, water, and ground-surface concentrations of radionuclides at appropriate locations and times based on known or assumed meteorological and hydro- logical conditions. These quantities are then used to calculate the absorbed dose to individuals based on their locations relative to these radionuclide concentrations. To estimate internal dose, the biokinetic models described in Appen- dix I are used to estimate the fate of radionuclides that are taken into the body by inhalation, ingestion, or absorption through skin. Radiation doses from internally deposited radionuclides are estimated by determining the spatial and temporal distribution of energy deposited in tissues and organs as a result of radioactive decay. Generally, this requires knowledge of the distribution of sources and targets in space and time. The source is the radionuclide of concern in a particular organ, tissue, or route of transit in the body. The target is the biological entity considered most relevant for determining dose and risk, which can range from molecules and cells for microdosimetry models to organs, tissues, or whole organisms. For radia- tion protection and epidemiologic studies, the level of averaging of radia- tion doses has consistently been at the tissue or organ level. Retrospective dose assessments related to effluent releases of radioac- tive materials into the environment can be classified in two categories: 1. The dose assessments made for establishing compliance with stan- dards or regulations. Usually, the calculated dose is much lower than the dose limit or standard. Under those conditions, the ratio- nale is to show that the calculated dose is an overestimate. Upper bound values of parameters such as the time spent at the location of maximum exposure or the consumption rates of local foodstuffs are used to demonstrate that there is no doubt that the calculated doses are below the dose limits or standards and, therefore, that there is no need to evaluate the uncertainties in the calculated doses.
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103 RADIATION DOSE ASSESSMENT • he calculated doses are expressed in terms of equivalent dose T (for specific organs or tissues) or effective dose (to take into account the irradiation of all organs of the body) in rem or in sievert because the dose limits or standards are expressed in those quantities. • he equivalent dose per unit intake (for internal irradiation) or T per unit exposure (for external irradiation) is the product of the absorbed dose per unit intake or exposure, which is a physical quantity, and a factor representing the biological effectiveness of the type of radiation that is considered. The value of this fac- tor, called the “radiation-weighting factor” and denoted as wR in ICRP Publications 60 and 103 (ICRP, 1991, 2007b), is based on experimental data for the relative biological effectiveness of various types of radiations at low doses, biophysical consider- ations, and expert judgment. The values for equivalent dose per unit intake and equivalent dose per unit exposure are set by the regulatory agency and, by convention, have no uncertainty. • he dose limits or standards apply to equivalent doses due to 1 T year of effluent releases. In the case of intakes of radionuclides with long biological times of residence in the body, such as strontium-90 or plutonium-239, the equivalent doses are still delivered many years after the year of intake. These “commit- ted” equivalent doses are calculated for the entire period of time between the age at intake and age 70 and are not broken down on an annual basis. • he dose limits or standards apply to the sum of the equivalent T doses from all types of radiation. This means that the equivalent dose from high-LET (linear energy transfer) radiation, such as alpha particles, are not separated from the equivalent doses from low-LET radiations, such as photons and electrons. 2. Dose assessments made for research purposes, for example, in epidemiologic studies. For this application, the doses need to be calculated as realistically as possible and the uncertainties in dose estimates have to be evaluated. The dose estimates should have no bias (that is, they should not be overestimates or underestimates), implying that all parameter values should be chosen accordingly. This is particularly difficult when absorbed doses to specified indi- viduals have to be calculated, but no interviews to those persons are feasible, thus precluding the knowledge of their lifestyle and dietary habits. • he calculated doses are expressed in terms of absorbed doses T
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104 ANALYSIS OF CANCER RISKS to specific organs or tissues. The special name of the unit of ab- sorbed dose is the gray, which is equal to 100 rad (see Table 3.1). • he absorbed doses per unit intake (for internal irradiation) or T per unit exposure (for external irradiation) are physical quanti- ties. Their values may be adjusted to the individuals that are considered if there is justification for such adjustments. In fact, the absorbed doses per unit intake or exposure are often derived from the values recommended by the ICRP. • he absorbed doses are calculated on an annual basis for each T year of exposure, for example, from radioactive effluent releases. This means that in the case of intakes of radionuclides with long biological times of residence in the body, such as strontium-90 and plutonium-239, the absorbed doses must be calculated start- ing with the year of initial exposure and for each year afterward. • he annual absorbed doses must be calculated separately for the T low-LET and the high-LET radiation. The focus of this report is on the second category of retrospective dose assessment. 3.2 REPORTED RADIATION DOSES AROUND NUCLEAR PLANTS Reported radiation levels outside the property lines of nuclear plants are now (and have been in the past) low compared to natural background radiation exposure levels (see Section 3.5.1), which varies from plant to plant. Annual absorbed doses from naturally occurring terrestrial gamma sources and cosmic rays typically range from 50 to 100 millirad per year (mrad/yr) (free-in-air6). However, an individual living in close proximity to the property line (i.e., “fence line”) of a nuclear plant might receive slightly elevated annual doses. Even during periods when nuclear plants released orders of magnitude more activity on average than currently (see Chapter 2), estimated external radiation doses to even the most exposed individual as a result of plant airborne effluent releases was likely only a fraction of the dose received from ambient natural background radiation. TLD measurements at various locations at some nuclear plants suggest that the direct radiation dose from stored waste onsite and nitrogen-16 gamma rays (see Chapter 2) could have amounted to a significant fraction of the ambient natural background exposure level at plant fence lines. In fact, these exposures could have accounted for most of the dose to the MEI at these plants. However, the dose from direct radiation from stored waste and nitrogen-16 decreases rapidly with distance from the fence line 6 That is, uncorrected for shielding by housing and indoor radiation sources.
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105 RADIATION DOSE ASSESSMENT and is generally an insignificant contributor to population exposures. For example, conservative estimates of doses from nitrogen-16 and stored waste at the Dresden plant (located in Illinois) were reported to result in an an- nual dose on the order of 8 mR/yr in 2009 to the MEI who was assumed to live in a home at the plant fence line and fish outdoors in an unshielded area for several hours per day (Exelon, 2010). Most nuclear plant licensees use conservative assumptions in calculat- ing annual doses to MEIs. For instance, some licensees assume that all effluent releases occur at ground level even though most airborne releases are made from tall stacks. This conservative assumption results in estimated maximum offsite dose levels that are much higher than would actually occur at any offsite location, particularly when averaged over a calendar quarter or year. Nevertheless, in recent years the estimated MEI doses are mostly less than 1 mrem/yr (Daugherty and Conatser, 2008), small frac- tions of ambient natural background radiation dose levels. However, doses in the 1970s and 1980s at some nuclear plants were higher, but even these doses were still much lower than doses from natural background radiation. Table 3.2 compares estimates of MEI doses for the early years of reactor operations at selected nuclear plants with estimates for more recent years. The reported MEI doses shown in Table 3.2 are also generally con- sistent with independent measurements made at some of these sites. For example, the U.S. Department of Energy’s Environmental Measurements Laboratory measured the integrated exposure from airborne radioactivity at a location 1.3 km from the Millstone-1 plant (a boiling-water reactor [BWR]) over a period of 500 days in 1973-1974 (Beck, 1975; Gogolak and Miller, 1974a,b). The absorbed dose in air was 3.5 mrad (0.035 mGy), in TABLE 3.2 Comparison of Estimated Whole-Body Doses to the MEI from Effluent Releases and Direct Radiation from Selected Nuclear Plants Whole-Body Dose CED to MEI Plant (source) (mrem/year) Dresden (noble gases) 14 (1975) 0.9 (2009) 1.0 × 10–4 (2009) Dresden (liquid) 0.1 (1975) Dresden (direct) — 8.4 (2009) Oyster Creek (air) 0.0036 (2008) Oyster Creek (water) NA NA Millstone (air) 16 (1975) 0.33 (2010) Millstone (liquid) 0.2 (1975) 0.0012 (2010) Millstone (direct) (incl. in air dose) 0.19 (2010) North Anna (air) 1.3 (1984) 0.013 (2008) North Anna (liquid) 4.0 (1984) 0.36 (2008) NOTE: CED, committed effective dose; NA, not available. SOURCE: Compiled from facility Radiological Environmental Monitoring Program reports.
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106 ANALYSIS OF CANCER RISKS reasonable agreement with what would be expected based on reported ef- fluent releases over that time period, which ranged from 6 to 100 millicuries per second (mCi/s); the free-in-air natural terrestrial background radiation exposure at that site over the same period was 109 mrad. Comparisons of calculated and measured airborne exposures for other locations around the Millstone plant are shown in Table 3.3. The Health and Safety Laboratory (now the Environmental Measure- ments Laboratory) also made similar measurements at a second BWR plant (Oyster Creek) over a period of several months in 1972. The maximum estimated offsite annual absorbed dose in air ranged from 10 to 15 mrad close-in with measurable levels out to 7 miles (~11 km) (Harold Beck, per- sonal communication, unpublished). The U.S. Environmental Protection Agency (USEPA) made similar mea- surements near several plant sites in the 1970s (Kahn et al., 1970, 1971, 1974). Measurements at the Prairie Island plant (a pressurized-water reac- tor [PWR] located in Minnesota) indicated a whole-body dose to the MEI of about 0.6 mrem/yr, excluding carbon-14. USEPA measurements at the Haddam Neck plant (a PWR located in Connecticut) in 1974 indicated a maximum annual dose of 0.9 mrem. Based on measurements at the Dresden plant in 1968, USEPA estimated a maximum annual dose of 14 ± 5 mrem. The total noble gas releases to the atmosphere during 1968 for Dresden were about 6 petabecquerels (PBq = 1015 Bq), comparable to the releases for 1975 when the facility estimated (conservatively) a dose to the MEI from noble gases of 14 mrem/yr. As indicated in Chapter 2, the releases of carbon-14 are, as of 2010, included in the effluent release reports that are submitted by facility licens- ees. Table 3.4 provides the estimated carbon-14 releases and corresponding equivalent doses for a sample of reactors that supplied that information in TABLE 3.3 Measured and Calculated Airborne Exposures at Seven Locations near the Millstone Plant Measured Calculated Location Length of Monitoring Period Absorbed Absorbed Distance (km) and (August 1973 through March 1974) Dose in Air Dose in Air Compass Direction (hours) (mrad) (mrad) 1.3 NNE 4727 0.312 0.342 2.6 ENE 4832 0.403 0.448 4.6 NNE 4254 0.080 0.100 4.6 E 4216 0.126 0.217 5.2 NE 4511 0.181 0.176 6.8 NNE 2919 0.046 0.055 8.0 ENE 4806 0.144 0.152 SOURCE: Gogolak and Miller (1974b).
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107 RADIATION DOSE ASSESSMENT TABLE 3.4 Carbon-14 Atmospheric Releases (Ci) and Equivalent Doses to MEI (mrem) Reported in Selected 2010 Annual Radioactive Effluents Releases Reports (ARERR) Bone Total-Body C-14 Equivalent Equivalent Release Fraction Estimation Dose to Dose to MEIa (mrem) MEIb (mrem) Reactor Name (Ci) as CO2 Method BWR Brunswick 21 1 FSAR 2.4 (99%) 0.47 Cooper 11.6 5.1 Ci/GWth-y 1.52 (99%) Dresden 20 5.1 Ci/GWth-y 0.73 Grand Gulf 9.5 0.95 FSAR 5.94 (94%) Nine Mile Point 9.16 0.95 5.1 Ci/GWth-y 0.22 0.043 Pilgrim 8.54 0.99 Neutronic 0.089 (80%) 0.018 (60%) calculation Susquehanna 24.5 1 EPRI (2010) 6.45 (96%) 1.29 PWR Beaver Valley 22 0.4 3.9 Ci/GWth-y 5.63 (95%) Catawba 20.4 0.2 9.4 Ci/GWe-y 4.78 (100%) Diablo Canyon 22.3 0.3 3.4-3.9 Ci/GWth-y 0.37 (98%) NUREG (1979)c H.B. Robinson 5.04 0.26 (76%) 0.052 (96%) McGuire 20.2 0.2 9.4 Ci/GWe-y 0.92 (98%) 0.44 (67%) North Anna 17 0.3 EPRI (2010) 1.26 (98%) Palisades 7.69 0.3 Neutronic 0.10 0.021 calculation San Onofre 21.9 0.78 (90%) Sequoyah 19.2 0.2 3.9 Ci/GWth-y 1.94 (96%) Waterford 19.2 0.2 FSAR 3.8 (98%) Wolf Creek 8.8 0.3 EPRI (2010) 1.3 0.26 NOTE: EPRI, Electric Power Research Institute; FSAR, Final Safety Analysis Report. aThe figure given in parentheses represents the percentage of the maximum organ equivalent dose from atmospheric effluent releases that is due to C-14. bThe figure given in parentheses represents the percentage of the total body equivalent dose from atmospheric effluent releases that is due to C-14. cUSNRC (1979). their 2010 reports. Even though different assumptions were used by the facility operators to estimate both the releases and the equivalent doses, it is clear that carbon-14 is currently a major contributor to the equivalent dose to the MEI from atmospheric effluent releases. Not included in these estimates is the equivalent dose to the MEI from nitrogen-16 and stored wastes, which is, for some reactors, the most important contributor to the total equivalent dose to the MEI. Pacific Northwest Laboratory (PNL)7 has published estimates of col- 7 PNL was renamed as the Pacific Northwest National Laboratory in 1995. This laboratory is located in Richland, Washington, adjacent to the Hanford Site.
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132 ANALYSIS OF CANCER RISKS FIGURE 3.8 Variation in annual terrestrial free-in-air terrestrial and cosmic-ray natural background doses for selected facilities. SOURCE: TLD data from 2008- 2009 radiological environmental monitoring reports for the plants shown in the Figure 3.8.eps figure. bitmap FIGURE 3.9 Variations in background radiation around the Millstone plant for 2009 based on TLD data. Note the relatively higher values near the fence line and variations with distance and direction. SOURCE: Dominion Nuclear Connecticut, Figure 3.9.eps Inc. (2009). bitmap
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133 RADIATION DOSE ASSESSMENT 3.5.2 Other Sources of Radiation Individuals living near nuclear facilities receive radiation from a num- ber of other sources besides background radiation. Arguably, depending on age and lifestyle factors, the two largest of these may be radiation from medical diagnostic17 procedures and air travel. These sources and their impacts on epidemiologic studies are described briefly in this section. The NCRP estimates that the average person in the United States is ex- posed to almost as much radiation from medical procedures each year (~3 mSv annual effective dose) as from background radiation including radon (~3.1 mSv annual effective dose) (NCRP, 2009a). Radiation from medical procedures has increased more than seven times since the 1980s when the last NCRP report was published (NCRP, 1987), whereas radiation from natural background sources has remained unchanged. The most significant changes in medical imaging were attributed to rapid increases in usage of computed tomography (CT) and nuclear medicine procedures. The exposures of particular individuals could be higher or lower than these averages depending on how many medical diagnostic procedures that use radiation they receive in any given year. There is no way to determine an individual’s exposure to medical radiation without interviewing them, but even in these cases there are likely to be large uncertainties in estimated exposures. These uncertainties arise from recall bias (i.e., the individual’s ability to recall the number, type, and dates of procedures) as well as the large variation in radiation doses that an individual receives from a given medical procedure depending, for example, on that individual’s age and what body part is being irradiated. Medical radiation could be a potential confounding factor in an epide- miologic study if individuals who live closer to nuclear facilities are exposed to radiation from medical diagnostic procedures at different rates compared to those who live farther away. This differential exposure could be due, for example, to differences in access to health care based on socioeconomic status. Confounding from medical radiation is likely to be less of a concern in epidemiologic studies that focus on children because they are less likely than adults to have received medical procedures involving high doses of radiation (e.g., CT scans, cardiac nuclear medicine procedures), although in utero exposure may be of concern (see, e.g., Table 3.14 in NCRP, 2009a). Air travelers are also exposed to increased levels of radiation resulting from galactic cosmic radiation.18 This radiation is primarily energetic pro- 17 Exposure to radiation from radiation therapy is not discussed here. About 1 percent of individuals having diagnostic procedures are believed to be undergoing radiotherapy. The doses from radiotherapy are on the order of 5,000 to 50,000 times as large as diagnostic procedures (NCRP, 2009). 18 Solar disturbances (e.g., solar flares) can also inject energetic particles into the Earth’s atmosphere.
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134 ANALYSIS OF CANCER RISKS tons (i.e., hydrogen nuclei) and alpha particles (i.e., helium nuclei). These particles interact with air molecules in the atmosphere and generate ad- ditional ionizing radiations including neutrons, protons, muons, electrons/ positrons, and photons. In general, the amount of radiation received during any particular flight depends on its altitude, latitude, and duration.19 For example, a 13-hour one-way flight from New York to Tokyo flown at a maximum altitude of 43,000 feet is estimated to result in an effective dose of about 0.0754 mSv (i.e., 7.54 mrem).20 Radiation from air travel could be a risk factor in epidemiologic stud- ies involving individuals who are frequent air travelers. There is no way to determine an individual’s exposure to radiation from air travel without interviewing them, but even in these cases there is likely to be large uncer- tainties in estimated exposures owing to recall bias. Exposure due to air travel is likely to be less of a concern in epidemiologic studies that focus on children because they are less likely than adults to have undertaken extensive air travel. 3.5.3 Exposures to Other Hazardous Materials Exposure to other hazardous materials, most notably toxic chemicals released from industrial facilities, can lead to a number of health outcomes including cancer (IARC, 2011; DHHS, 2011). Many of the front-end nu- clear facilities discussed in Section 3.2 also release chemicals. Furthermore, it is well known that the chemical toxicity of some radioactive effluents such as uranium may be more deleterious than the low levels of radio- activity (Bleise et al., 2003). Consequently, chemical exposures could be an important risk factor in epidemiologic studies of populations that are exposed to both radiation and chemical hazards. This could be especially problematic if the epidemiologic study focuses on cancers that have both radiation and chemical etiologies such as bladder cancer and leukemia. It will be important to identify major industrial facilities in the vicin- ity of nuclear facilities that are examined in the epidemiologic study. For example, the Metropolis, Illinois, conversion facility discussed earlier is co-located with a large chemical plant. The annual material releases from industrial facilities can be obtained from the USEPA21 and assessed to determine their potential impact on the epidemiologic study. It might be 19 The Earth’s atmosphere and magnetic fields shield this radiation. As a consequence, less radiation is received at lower altitudes and at locations closer to the Earth’s equator. 20 See http://www.faa.gov/library/reports/medical/oamtechreports/2000s/media/0316.pdf. 21 USEPA’s Toxics Release Inventory program (see www.epa.gov/tri/) maintains a database on releases of over 600 toxic chemicals from facilities in the United States. Facility owners are required to provide information on their toxic releases to USEPA on an annual basis. The database was complete through 2010 when the present report was in development.
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135 RADIATION DOSE ASSESSMENT necessary to exclude particular census tracts or cancer types from the epi- demiologic study in cases where there are substantial industrial releases. This will need to be handled on a facility-by-facility basis. 3.6 CHARACTERIZING AND COMMUNICATING UNCERTAINTIES The uncertainties in dose estimates for an epidemiologic study are likely to be substantial. These uncertainties arise from uncertainties in source terms (i.e., reported effluent releases; see Chapter 2) and, usually to a greater extent, uncertainties in atmospheric transport and liquid dispersion models that relate these source terms to environmental concentrations, and also uncertainties in pathway models that relate environmental concentra- tions to dose. Uncertainties in dose estimates have the potential to mask the “true” dose-response relationship in an epidemiologic study. Consequently, understanding and characterizing these uncertainties is important. The magnitude of dose estimate uncertainties is also likely to vary over time. Effluent release data for early years of facilities operations are of lower quality than more recent data (see Chapter 2). As a consequence, dose estimates based on earlier data are likely to be more uncertain than doses calculated for releases for more recent years. Moreover, because efflu- ent releases in earlier years were much higher as a result of higher airborne effluent releases (see Chapter 2), uncertainties in airborne effluent releases are likely to be relatively more important than uncertainties in liquid ef- fluent releases. The airborne effluent release uncertainties are a function of how representative the weekly grab samples22 were with respect to the ac- tual releases of specific nuclides, as well as to uncertainties in stack airflow rates, especially if they varied with time. There is much less uncertainty associated with the measured activities of the grab samples themselves. Furthermore, the use of an average quarterly value for batch releases rather than the actual values for each batch adds to the reported uncertainties and resultant dose estimates, particularly for PWRs. Uncertainties in diffusion and dispersion models that relate source terms (effluent releases) to environmental concentrations as well as expo- sure pathway models relating environmental concentrations to doses can be high. Atmospheric dispersion estimates can also be very uncertain, particu- larly when releases are episodic, when there are terrain irregularities, and for locations that are distant from the facility fence line (Table 3.14). On sites with flat terrain, Gaussian plume models have been shown to provide reasonable estimates of air concentrations when integrated over a sufficient 22 Effluent releases of specific radionuclides for continuous (as opposed to batch) releases are based on analyses of weekly grab samples rather than continuous monitoring. See Ap- pendix H.
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136 ANALYSIS OF CANCER RISKS TABLE 3.14 Uncertainties in Gaussian Plume Models Range, Predicted over Observed Conditions Air Concentration (P/O) Highly instrumented site; ground-level, centerline; within 0.65 to 1.35 10 km of a continuous point source Specific time and location, flat terrain, steady 0.1 to 10 meteorology, within 10 km of release point Annual average, specific location, flat terrain, within 0.5 to 2 10 km of release point Annual average, specific location, flat terrain, 10 m to 0.25 top 4 150 km downwind Complex terrain or meteorology, episodic releases 0.01 to 100 Episodic, surface-level releases, wind speeds less than 1 to 100 2 m s –1 SOURCE: Miller (1995). time interval, although estimates for a shorter integration times can be very uncertain. Uncertainties increase for sites with complex terrain (e.g., sites with hills or valleys). Also, local meteorology at any particular time (wind speed, direction, and atmospheric stability) can vary significantly from an- nual averages and result in significant errors if the latter are used to estimate doses for batch effluent releases into the atmosphere. Liquid diffusion models for effluent releases into estuaries, lakes, and oceans, as well as spills into surface and ground water, are very crude. Ad- ditionally, estimates of environmental usage of potentially contaminated water are also very crude in the absence of subject interviews. Thus, most estimated doses resulting from liquid effluents to representative individuals residing in specific locations are likely to be highly uncertain and will vary significantly from individual to individual and location to location. As discussed in Chapter 2, effluent emissions varied widely over time and generally decreased rapidly with distance from the facility fence line. Exposed persons were not at the same place with respect to the facility at all times. Consequently, the dose to any particular individual will be even more uncertain than the dose to an unspecified individual at a particular location and time. For studies that are based on individuals (such as a cohort or a case-control study) that require individual dosimetry data, this uncertainty will depend on the ability to determine individual lifestyle behaviors. Considering the complexity and range of uncertainties discussed above, a detailed quantitative analysis of uncertainty in an epidemiologic study is not practical, particularly for an ecologic study. An extensive quantitative analysis would require resources and effort not commensurate with the magnitude of the likely doses, the quality of the effluent release data, and the degree of complexity recommended by the committee for dose recon- struction. However, a quantitative or at least semiquantitative uncertainty
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137 RADIATION DOSE ASSESSMENT analysis could be performed, at least for a few facilities and years of opera- tion, for the case-control study. Nevertheless, at the very least, any epidemiologic study will need to address uncertainty, at least qualitatively. Such an analysis should: • Identify, evaluate, and rank all potential sources of major uncer- tainty and identify site-to-site and temporal differences; • Identify potential bias versus random errors in the dose calculations that could affect interpretation of the epidemiologic findings; and Identify shared errors23 as opposed to stochastic variability to • properly evaluate the risk from radiation exposure should any increased risk of cancer be identified. Although the reported environmental monitoring data for almost all sites and times was either below minimum detectable levels or, for external radiation, not distinguishable from background, an epidemiologic study could still use these data to set upper limits on the reported effluents by back-calculating from the minimum detection levels. This would at least place upper bounds on effluent releases. 3.7 FINDINGS AND RECOMMENDATIONS This chapter provides the committee’s assessment of methodological approaches for assessing offsite radiation doses to populations living near nuclear plants and fuel-cycle facilities to support an epidemiologic study. Based on this assessment, the committee finds that: 1. Absorbed dose—the energy deposited by ionizing radiation per unit mass of tissue in specific organs of interest—is the appropriate dose quantity for use in an epidemiologic study. Other dose quantities, for example effective dose, equivalent dose, and collective dose, are designed for regulatory purposes and are not appropriate for epidemiologic studies (see Section 3.4.1). The dose to a maximally exposed individual (MEI) is also not an appropriate quantity for an epidemiologic study because it provides a high-sided estimate at 23 A sdiscussed in NCRP (2009b), uncertainties that are common to many individuals (for example, error in the amount of effluents from a facility) can introduce bias (systematic uncer- tainty) in estimated doses compared to uncertainties that are unshared and represent stochastic variability in true doses among individuals. When uncertainties are shared among individuals in a population, the degree of variability in true doses among individuals is less than would be estimated by assuming that uncertainties in each individual’s dose are purely random. An overestimation of the variability in true doses among individuals results in a suppression of dose-response relationships derived in an epidemiologic study, i.e., the true dose response is flattened (Schafer and Gilbert, 2006).
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138 ANALYSIS OF CANCER RISKS a single spatial point and does not reflect the variation is dose with distance and direction from a nuclear facility. 2. Absorbed doses to individuals attributable to living near nuclear plants and fuel-cycle facilities are anticipated to be very low (see Sections 3.2 and 3.3), in most cases well below variations in lev- els of natural background radiation in the vicinity of the facility and from facility to facility. These doses are also anticipated to be lower than levels of radiation received by some members of the public from medical procedures and air travel. Consequently, dose estimates used in an epidemiologic study need to account for these other radiation exposures and other risk factors such as exposure to hazardous (and potentially carcinogenic) materials released from industrial facilities located near nuclear facilities (see Section 3.5). 3. Estimates of doses to individuals living around nuclear facilities will have uncertainties owing to facility effluent releases, dose models, and other risk factors. A detailed quantitative analysis of uncertainty is not practical. However, a qualitative uncertainty analysis can be performed for a few facilities and years of operation to estimate the probably magnitudes of these uncertainties (see Sec- tion 3.6). It will be important to communicate these uncertainties to stakeholders as part of the epidemiologic study. 4. Computer models have been developed to estimate absorbed doses in individuals exposed to radiation through environmental path- ways. These existing models could be adapted or a new model could be developed to estimate doses to individuals living near nuclear facilities to support an epidemiologic study. Regardless of the approach used, it is essential that the underlying computer model reflect modern practices for dose reconstruction (see Section 3.4). In light of these findings, the committee recommends that a pilot study be undertaken to demonstrate the feasibility of reconstructing absorbed doses for an epidemiologic study. This pilot study should: 1. Develop a computer model (i.e., by modifying or adapting an existing model or building a new model) to obtain estimates of absorbed doses to the whole body and individual organs result- ing from airborne and waterborne effluent releases. This model should be similar in scope and complexity24 to that used by the 24 The committee uses the phase “similar in scope and complexity” to mean that the model should use the same general approach as the PNL model to estimate annual absorbed doses as a function of direction and distance from a facility based on effluent release and meteorological data averaged over daily to quarterly periods.
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139 RADIATION DOSE ASSESSMENT Pacific Northwest Laboratory (Baker, 1996) to estimate doses to populations living near nuclear plants in the 1970s and 1980s, but it should be updated as described in Section 3.4 to provide point and census-tract estimates of absorbed dose using modern dose reconstruction practices. 2. Demonstrate the utility of this model for dose reconstruction to support the epidemiologic study designs recommended in Chapter 4 (See Section 4.4 in Chapter 4) by: • sing the model to obtain dose estimates as a function of dis- U tance (0 to 50 kilometers [30 miles] from the plant) and direction for the six nuclear plants and one fuel-cycle facility subject to the pilot study in Chapter 2 (see Chapter 2, Section 2.5). • eveloping a methodology to account for natural background D radiation and, to the extent feasible, other sources of radiation in the dose estimates. • ndertaking an uncertainty analysis as described in Section 3.6. U The results of this pilot study should be used to inform decisions about any Phase 2 epidemiologic study effort. REFERENCES Baker, D. A. (1996). Dose Commitments due to Radioactive Releases from Nuclear Power Plant Sites: Methodology and Data Base. NUREG/CR-2850 (PNNL-11190), Supp. 1. Beck, H. L. (1975). Techniques for Monitoring External Environmental Radiation around Nuclear Facilities. Proceedings of the 8th Annual Conference On Nuclear Safety Research (in Japanese) (May). Beck, H. L., and K. M. Miller (1982). Temporal Variations of the Natural Radiation Field. Trans. of Second Special Symp. on the Natural Radiation Environment, Wiley Eastern. Bleise, A., P. Danesi, and W. Burkart (2003). Properties, use and health effects of uranium. J. Environ. Radioact. 64:93-112. BNL (Brookhaven National Laboratory) (1979). Radioactive Materials Released from Nuclear Power Plants, 1977, NUREG-0521 (January). Commonwealth Edison (1976). Semi-annual Report Pertaining to Radioactive Effluent Dis- charges, Environmental Monitoring, Solid Radioactive Waste, and Personnel Exposures for Dresden Units 1, 2, and 3 for the Time Period July 1, 1975 through December 31, 1975 (February). Crow Butte Resources, Inc. (2010). Uranium Project Radiological Effluent and Environmental Monitoring Report for Third and Fourth Quarters, 2010. EPRI (Electric Power Research Institute] (2010) Estimation of Carbon-14 in Nuclear Power Plant Gaseous Effluents, Report 1021106 (December 23). Daugherty, N., and R. Conaster. (2008). Radioactive Effluents from Nuclear Plants: Annual Report 2008. Washington, DC: Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission. DHHS (U.S. Department of Health and Human Services) (2011). Public Health Service, Na- tional Toxicology Program. Report on Carcinogens, 12th Edition. Available at http://ntp. niehs.nih.gov/ntp/roc/twelfth/roc12.pdf.
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142 ANALYSIS OF CANCER RISKS USNRC (1981). 40 CFR 190 Compliance Assessment for NRC Licensed Uranium Recovery Facilities as of December 1, 1980. Division of Waste Management, Uranium Recovery Licensing Branch. USEC (United States Enrichment Corporation) (2006). PGDP Quarterly Radiological Dis- charge Monitoring Report-Fourth Quarter 2006, Portsmouth Quarterly Radiological Discharge Monitoring Report-Fourth Quarter 2006. USEC (2008). Paducah Gaseous Diffusion Plant, Docket No. 70-7001. Application for Re- newal of Certificate of Compliance GDP-1, April 10. Westinghouse (2002). Westinghouse Electric Company LlC Nuclear Fuel, Columbia Plant ALARA Report, Calendar Year 2002.