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Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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3

Radiation Dose Assessment

This 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 effluent 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

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

TABLE 3.1 Selected Quantities and Units for Radiation Exposure and Dose

Quantity Old Unit SI Unit or
Its Special
Name
Between Units
Relationship
Field of
Application
Reference
Exposure R C kg-1 1 R = 2.58 10-4
C kg-1
Monitoring NCRP
(2007)
Absorbed dose rad Gy 1 rad = 0.01 Gy Research ICRP
(2007b)
Dose equivalenta rem Sv 1 rem = 0.01 Sv Radiation Protection ICRP
(1977)
Equivalent dosea rem Sv 1 rem = 0.01 Sv Radiation protection ICRP
(1991)
Effective dose
equivalentb
rem Sv 1 rem = 0.01 Sv Radiation protection ICRP
(1991)
Effective doseb rem Sv 1 rem = 0.01 Sv Radiation protection ICRP
(1991)
Committed effective
dose equivalentc (CEDE)
rem Sv 1 rem = 0.01 Sv Radiation protection ICRP
(1991)
Collective dose
equivalent
person-rem person-Sv 1 person-rem =
0.01 person-Sv
Radiation protection ICRP
(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 (wR) for the calculation of equivalent doses.

bEffective dose equivalent and effective dose are conceptually similar. Effective dose equivalent 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 approximately the same value for all organs and tissues of the body, as is the

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

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 exposed 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 Commission 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 application 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:

  • prospective dose assessment for planning and optimization of protection; and
  • retrospective dose assessment for demonstrating compliance with dose 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 radioactive 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.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

purposes worldwide, and the latter is used in the current USNRC dose compliance 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 example, 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 exposure5 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 radioactive 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” instead 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.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×
  • 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 radiation, 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

image

FIGURE 3.1 Pathways for exposure to radiation from effluent releases from nuclear plants and fuel-cycle facilities. SOURCE: Soldat et al. (1974).

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×
  • from a facility, from radionuclides present in air, or from radionuclides 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 external 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 Appendix 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 radiation protection and epidemiologic studies, the level of averaging of radiation doses has consistently been at the tissue or organ level.

Retrospective dose assessments related to effluent releases of radioactive materials into the environment can be classified in two categories:

1. The dose assessments made for establishing compliance with standards or regulations. Usually, the calculated dose is much lower than the dose limit or standard. Under those conditions, the rationale 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.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×
  • The calculated doses are expressed in terms of equivalent dose (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.
  • The equivalent dose per unit intake (for internal irradiation) or 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 factor, 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 considerations, 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.
  • The dose limits or standards apply to equivalent doses due to 1 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 “committed” 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.
  • The dose limits or standards apply to the sum of the equivalent 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 individuals have to be calculated, but no interviews to those persons are feasible, thus precluding the knowledge of their lifestyle and dietary habits.

  • The calculated doses are expressed in terms of absorbed doses
Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×
  • to specific organs or tissues. The special name of the unit of absorbed dose is the gray, which is equal to 100 rad (see Table 3.1).
  • The absorbed doses per unit intake (for internal irradiation) or per unit exposure (for external irradiation) are physical quantities. 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.
  • The absorbed doses are calculated on an annual basis for each 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 starting with the year of initial exposure and for each year afterward.
  • The annual absorbed doses must be calculated separately for the 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.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

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 annual 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 calculating 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 fractions 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 consistent 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 MEIfrom Effluent Releases and Direct Radiation from Selected Nuclear Plants

Plant (source) Whole-Body Dose CED to MEI (mrem/year)

Dresden (noble gases) 14 (1975) 0.9 (2009)
Dresden (liquid) 0.1 (1975) 1.0 * 10-4 (2009)
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.
Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

reasonable agreement with what would be expected based on reported effluent 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 Measurements 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, personal communication, unpublished).

The U.S. Environmental Protection Agency (USEPA) made similar measurements near several plant sites in the 1970s (Kahn et al., 1970, 1971, 1974). Measurements at the Prairie Island plant (a pressurized-water reactor [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 licensees. 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 SevenLocations near the Millstone Plant

Location
Distance (km) and Compass Direction
Length of Monitoring Period
(August 1973 through March 1974)
(hours)
Measured Absorbed Dose in Air
(mrad)
Calculated Absorbed Dose in Air
(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).
Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

TABLE 3.4 Carbon-14 Atmospheric Releases (Ci) and Equivalent Doses to MEI (mrem) Reported in Selected 2010 Annual Radioactive Effluents Releases Reports (ARERR)

Reactor Name C-14
Release
(Ci)
Fraction
as CO2
Estimation
Method
Bone
Equivalent
Dose to
MEIa(mrem)
Total-Body
Equivalent
Dose to
MEIb(mrem)
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 calculation 0.089 (80%) 0.018 (60%)
 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%)
 H.B. Robinson 5.04 NUREG (1979)c 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 calculation 0.10 0.021
 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 fgure given in parentheses represents the percentage of the maximum organ equivalent dose from atmospheric effuent releases that is due to C-14.

bThe fgure given in parentheses represents the percentage of the total body equivalent dose from atmospheric effuent 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 collective

7 PNL was renamed as the Pacific Northwest National Laboratory in 1995. This laboratory is located in Richland, Washington, adjacent to the Hanford Site.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

doses8 to populations living in the vicinity of operating nuclear plants in the United States resulting from airborne and waterborne effluent releases (NUREG/CR-28509). Figure 3.2 shows PNL’s collective dose estimates for persons living between 2 and 80 km from selected nuclear plants that have a range of effluent releases.10 As can be seen from the figure, the total collective doses for some plants (e.g., Millstone and Dresden plants) were several orders of magnitude higher than for other plants (e.g., Fort Calhoun and Trojan plants). The estimated collective doses generally correlate with total noble gas effluent releases from the plants. Note that most of the collective dose for each site was usually delivered in only a few years (but not necessarily the same years) as shown in Figure 3.3. The 12 nuclear plants with the largest effluent releases accounted for over 75 percent of the total collective doses from all nuclear plants. Nuclear plants that have had high and low collective dose impacts over their operating histories are listed in Table 3.5.

Because the calculated collective doses are integrals over 2-80 km, they do not reflect the dose to MEIs or to populations living within 2 km of the plants. In addition, neither the doses resulting from atmospheric releases of carbon-14 nor the doses incurred prior to 1975 are included in the estimates shown in the table. Based on reported total effluent releases, the additional collective dose from operations prior to 1975 may have been comparable or greater at some plants, and the collective dose from atmospheric releases of carbon-14 may be a more significant contributor to the collective dose in more recent years as releases from other radionuclides have decreased dramatically (see Section 2.1 in Chapter 2).

For illustrative purposes, Table 3.6 lists the radionuclides that were reported by facility operators to make the highest contributions to collective doses from effluent releases (airborne and waterborne) in 1988 from 71 operating commercial nuclear plants. The relative contributions of each radio-nuclide to the total collective doses from all 71 plants are also shown in the table. It is clear that, at least in 1988 and probably since that time, tritium (hydrogen-3) has played an important role, both for airborne and waterborne releases. For airborne releases, isotopes of noble gases (krypton-88, xenon-133, and xenon-135) also contributed substantially to the collective dose, whereas iodine-131 was not a critical radionuclide for any of the

8 These collective dose data are presented here because they are the only data the committee could find that provide some basis for comparing doses to populations living near different nuclear plants. As noted earlier in this chapter, collective dose is not an appropriate metric for epidemiologic studies.

9 PNL issued a series of reports entitled Dose Commitments Due to Radioactive Releases from Nuclear Power Plant Sites that covered nuclear plant operations from 1977 to 1992. The first four reports in the series were issued as PNL-2439 (1977), NUREG/CR-1125/PNL-2940 (1979), NUREG/CR-1498/PNL-3324 (1980), and NUREG/CR-2201/PNL-4039 (1982). The remaining reports were issued from 1982 to 1996 as NUREG/CR-2850, vols. 1-14.

10 As shown in Chapter 2, effluent releases among nuclear plants can vary substantially.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

image

FIGURE 3.2 Collective doses to populations living between 2 and 80 km from selected nuclear plants. SOURCE: NUREG/CR-2850 (PNL-4221), vol. 14.

image

FIGURE 3.3 Collective doses to populations living between 2 and 80 km from the Millstone, Dresden, and Oyster Creek plants, 1975-1992. SOURCE: NUREG/CR-2850 (PNL-4221), vol. 1-14.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

TABLE 3.5 Nuclear Plants with High and Low Collective Dose Impacts over their Operating Histories

Site (number and type of reactors) State Population within
5 Miles in 2000
(thousands)
Collective Dosea
1975-1992
(person-rem)
Maximum Annual
Collective Dose
(person-rem)
Maximum
Dose Year
Total Noble Gas Releases
1975-1992 (GBq)
Plants Having High Dose Impacts
 Millstone (1 BWR, 2 PWR) CT 53.3 1384 750 1975 1.80 × 108
 Dresden (BWR) IL 22.9 879 360 1975 1.20 × 108
 Oyster Creek (BWR) NJ 44.2 594 220 1979 9.40 × 107
 Browns Ferry (BWR) AL 6.1 232 106 1984 9.00 × 107
 Nine Mile Point. 2(BWR) NY 6.7 227 140 1979 5.50 × 107
 Zion (PWR) IL 177 34 1984 1.70 × 106
 McGmre (2PWR) NC 51.2 165 20 1984 1.10 × 10s
 Oconee (3PWR) SC 15.6 157 38 1977 1.50 × 107
 Peach Bottom (BWRs) PA 11.3 134 30 1979 3.20 × 107
 North Anna (2PWR) VA 6.9 123 44 1984 3.30 × 10s
 Quad Cities (2 BWR) IL 6.3 121 42 1980 1.20 × 107
 Pilgrim (BWR) MA 23.1 95 52 1977 3.10 × 107
 Cook (2 PWR) MI 17.0 90 40 1978
 Indian Point (1BWR, 2PWR) NY 88.2 85 13 1977 5.00 × 106
 Hatch (2 BWR) GA 2.1 73 35 1977 6.90 × 106
 LaCrosse (BWR) WI 59 12 1976 9.00 × 10s
 Big Rock (BWR) MI 57 10 1980 1.70 × 107
 Brunswick (2 BWR) NC 13.4 45 14 1982 8.20 × 107
 Crystal River (3 PWR) FL 6.1 44 19 1981 7.20 × 106
 Haddam Neck (PWR) CT 40 7.5 1980 2.30 × 106
 Arkansas (2 PWR) AK 14.2 39 4.7 1985 5.70 × 106
Plants Having Low Dose Impacts
 Fort Calhoun (PWR) NE 9.3 7.3 1.9 1984 6.0 × 105
 Duane Arnold (BWR) IA 12.2 2.7 0.87 1978 1.0 × 106
 Trojan (PWR) OR 1.7 0.25 1991 4.1 × 105
 Cooper (BWR) NE 0.9 1.6 0.39 1976 4.6 × 106

NOTE: BWR, boiling-water reactor; PWR, pressurized-water reactor.

aFor individuals living between 2 and 80 km of the plant boundary.

SOURCE: Population information from Table 1.3 in Chapter 1; other information from NUREG/CR-2850 (PNL-4221), vol. 14.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

TABLE 3.6 Radionuclides with the Highest Contribution to Collective Dose from Effluent Releases (Airborne and Waterborne) in 1988 from the 71 Operating Commercial Nuclear Plants

Airborne Releases Waterborne Releases
Radionuclide Number of Nuclear Power Plants
with the Highest Contribution to
Collective Dose
Relative Contribution
to the Total Collective
Dose (Percent)
Number of Nuclear Power Plants
with the Highest Contribution to
Collective Dose
Relative Contribution
to the Total Collective
Dose (Percent)
Tritium 39 28 18 35
Carbon-14a 1 0.4 0 0.1
Manganese-54 0 <0.1 1 2.8
Iron-55 0 0 4 0.5
Iron-59 0 0 1 0.2
Cobalt-58 0 <0.1 1 1.1
Cobalt-60 2 1.0 3 2.3
Zinc-65m 0 <0.1 4 1.6
Krypton-88 8 7 0 0
Strontium-90 1 <0.1 1 1.4
Iodine-131 0 <0.1 1 0
Xenon-133 17 31 0 0
Xenon-135 3 13 0 0
Cesium-134 0 <0.1 11 28
Cesium-137 0 0.3 18 24
No release 0 8
Total number of plants 71 71
Total collective dose 9.6 65
(person-rem)

aThe collective dose from releases of carbon-14 was calculated for only two power plants.

SOURCE: NUREG/CR-2850, vol.10.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

plants. With respect to waterborne releases, cesium-134 and cesium-137 were the two most important radionuclides, in addition to tritium.

It is worth noting that the collective dose from carbon-14 was apparently calculated for only two nuclear plants (Ginna and Yankee Rowe) and was found to be the highest contributor to collective dose from airborne releases for one of those (Yankee Rowe). Had the collective doses from carbon-14 releases been estimated and reported for the other nuclear plants, it is likely that it would have been found among the main contributors to the collective dose from airborne effluent releases, assuming that the results of Table 3.4 for the dose to the MEI can be translated in terms of collective dose.

Figure 3.4 shows the reported annual collective doses from airborne and waterborne radioactive effluent releases from all operating nuclear plants from 1975 to 1992. In the early years of operations when doses were highest, most of the collective dose was from exposure to airborne effluents. In contrast, most of the collective dose in recent years is from waterborne releases, but these collective doses remained fairly constant over time. The contribution to the collective dose from waterborne versus airborne releases differed at different sites depending on such factors as the presence of nearby recreational facilities (e.g., rivers and lakes).

External radiation exposures around nuclear plants would be expected to vary not only with distance from the plant site, but also with direction,

image

FIGURE 3.4 Collective doses from air and liquid effluents at all operating nuclear plants from 1975 to 1992. SOURCE: NUREG/CR-2850 (PNL-4221), vol. 14.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×



local topography, and stack height, particularly for sites where wind directions are not distributed isotropically. Consequently, spatial and directional variations in dose could be significant at some plant sites and could also vary with season. If so, the use of annual effluent releases and annual average meteorology to estimate doses would not reflect these spatial variations. This would be particularly true for plants that do not release effluents randomly in time such as PWRs, which release effluents in batches.

To illustrate, Figures 3.5 and 3.6 show the wind rose and calculated 1975 external doses around the Dresden plant. Both the wind rose and dose distributions display asymmetry. Residents living north of the plant received higher doses relative to residents living in other compass directions at a given distance from the plant site. It is likely that the asymmetry in calculated dose at some sites was even more pronounced.

With regard to waterborne releases, the degree of asymmetry is more difficult to predict. The degree of asymmetry depends to a large extent on the distribution of contaminated drinking water and contaminated foodstuffs (fish and invertebrates).

image

FIGURE 3.5 Annual wind rose for the Dresden plant for all stability classes and speeds combined at the height of the plant stack. The concentric lines indicate the percent time (from 0 to 7 percent) the wind was blowing. The radial lines show the compass direction that the wind was blowing. SOURCE: Commonwealth Edison (1976).

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

image

FIGURE 3.6 Calculated annual dose contours (rem) for 1975 at the Dresden plant from airborne effluent releases for comparison with the average wind rose (Figure 3.4). SOURCE: Commonwealth Edison (1976).

3.3 REPORTED DOSE ESTIMATES AROUND NUCLEAR FUEL-CYCLE FACILITIES

As is the case for nuclear plants (Section 3.2), doses to MEIs living near fuel-cycle facilities in recent years are very low. Some examples of MEI dose estimates for fuel-cycle facilities are shown below:

  • Milling (Crow Butte): 0.68 mrem/yr in 2010 (Crow Butte Resources, 2010).
  • Conversion (Honeywell): 0.57 mrem/yr in 2005 (Honeywell, 2006).
Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×
  • Enrichment (Paducah): 0.94 mrem/yr in 2009 (Portsmouth, 2009).
  • Fuel Fabrication (Nuclear Fuel Services): 0.002 mrem/yr in 2009 (NFS, 2009).

However, doses in early years of operation might have been significantly greater. The doses for various types of facilities are discussed below.

3.3.1 Mining and Milling Facilities

As noted in Chapter 1, the committee did not consider mining and milling facilities in this study because of their small surrounding populations (see Table 1.3 in Chapter 1). Because of the small populations, the collective doses to populations living within 80 km of these facilities have probably been small relative to collective doses to populations near nuclear plants. Doses in a recent year (2010) at a typical in situ uranium recovery facility (Crow Butte) to the MEI are estimated to be about 0.7 mrem/yr (0.5 mrem/yr from radon, the remainder from uranium). Doses in earlier years were much greater as shown in Table 3.7. External (direct radiation)

TABLE 3.7 Reported 50-Year Committed Doses to the MEI for 1979 or 1980 Effluent Releases from In Situ Uranium Recovery Facilities in the United States


Facility Location Whole-Body
(mrem)
Bone
(mrem)
Lung
(mrem)

Atlas Minerals Moab, UT 2.4 34.6 74.8
Bear Creek Uranium Co. Converse Co., WY 0.486 6.14 0.782
Exxon Minerals Highland Mill Converse Co., WY 0.847 12.2 13.9
Federal-American Partners Gas Hills, WY 0.649 17.4 35.9
Energy Fuels Nuclear White Mesa Blanding, UT 1.40 15.0 2.24
Minerals Exploration Co. Sweetwater Co., WY 0.0081 0.0831 0.038
Pathfinder Mines Gas Hills, WY 0.599 11.4 15.7
Pathfinder Mines Shirley Basin, WY 1.61 18.0 6.56
Petrotomics Company Shirley Basin, WY 0.696 9.75 9.58
Plateau Resources Shootering Canyon, UT 0.135 3.60 6.63
Rio Algom Humeca Mill La Sal, UT 0.528 11.0 23.5
Union Carbide Corp. Gas Hills, WY 0.97 12.5 1.81
United Nuclear Corp. Morton Ranch Converse Co., WY 0.08 0.34 0.28
Western Nuclear Inc., Split Rock Jeffrey City, WY 2.0 24.2 11.5

NOTE: Committed dose is the total dose that would be received by an individual during a specified period (usually the 50-year period) following the intake of a radioactive material. The doses do not include contributions from radon because the dose criteria in 40 CFR 190(Environmental Radiation Protection Standards for Nuclear Power Operations) do not apply to dose from radon and its short-lived decay products.SOURCE: USNRC (1981).
Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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whole-body doses result primarily from exposure to mill tailings. Bone and lung doses result from inhalation of airborne effluents. The United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR, 1982) estimated that organ doses from mining and milling operations were mainly from inhalation or airborne emissions of radon decay products, with additional contributions from uranium and thorium isotopes, radium-226, and lead-210. The highest doses were to the lung and bone.

3.3.2 Uranium Conversion Facilities

The only uranium conversion facility in the United States is the Honeywell plant, which is located at Metropolis, Illinois. The plant licensee estimated that the dose to the MEI in 2005 was 0.57 mrem (Honeywell, 2006). The MEI was located at the nearest residence, 564 meters (1,850 feet) north-northeast of the Metropolis facility. The MEI does not have a home garden; however, to be conservative, the ingestion pathway was included in the dose assessment. (The methodology, data, and assumptions used in the dose assessments were provided in Honeywell [2006]). Honeywell also estimated the annual collective dose to the population of about 517,000 people surrounding the facility as 0.0381 person-Sv (3.81 person-rem) per year.

The Paducah Gaseous Diffusion Plant is located near the Metropolis facility. Based on data reported by USEC, Inc., the radiation dose (TEDE11) to the MEI from atmospheric emissions from the Paducah Gaseous Diffusion Plant was estimated to be 3.54 × 10–4 mSv (0.0354 mrem) per year in 2004 (Honeywell, 2006). Therefore, the Paducah Gaseous Diffusion Plant would not contribute appreciably to the radiation dose for the Metropolis facility’s MEI.

Although the radiological impacts from current normal operations are very small, doses in early years of operation might have been greater. The committee did not investigate data on estimated doses from conversion for early years of operation.

3.3.3 Uranium Enrichment Facilities

The maximum dose that a member of the public was estimated to have received from reported effluent releases from the Portsmouth enrichment facility in 2009 was 0.94 mrem: 0.024 mrem from airborne radionuclides, 0.037 mrem from radionuclides released to the Scioto River, 0.72 mrem from direct radiation from the depleted uranium cylinder storage yards, and

11 Total effective dose equivalent. This is the sum of the effective dose equivalents from internal and external exposures.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

0.16 mrem from exposure to radionuclides detected at offsite monitoring locations (DOE, 2011). This dose calculation used a worst-case approach; that is, the calculation assumes that the same individual is exposed to the most extreme conditions from each pathway. The maximum potential doses in 2004 and 2005 were 1.86 mrem (DOE, 2006) and 1.67 mrem (USEC, 2006), respectively. The 2005 estimate broke down as follows: 0.012 mrem from airborne radionuclides, 0.025 mrem from radionuclides released to the Scioto River, 1.1 mrem from direct radiation, and 0.53 mrem from exposure to radionuclides detected at offsite monitoring locations. The relatively high external (direct) exposure is primarily from tanks of depleted uranium.

The maximum effective dose equivalent to the MEI for the Paducah plant was reported as 0.0433 mrem/yr in 2002 (USEC, 2008). Based on estimated 2002 census data, the total committed effective dose equivalent (CEDE) to the 50-mile population (approximately 531,000 persons, including 36,500 within 10 miles (~16 km) of the plant and approximately 104,000 within 20 miles [-32 km]) was image0.2 person-rem.

The committee did not attempt to find data for very early years of operation at these facilities.

3.3.4 Fuel Fabrication Facilities

The committee reviewed reported dose estimates for recent years for two currently operating fuel fabrication facilities: Nuclear Fuel Services, Inc. (Tennessee) and Westinghouse Electric Company, LLC Columbia Fuel Fabrication Facility (South Carolina).

Doses related to Nuclear Fuel Services (NFS) Erwin plant operations are dominated by airborne effluent releases. In 2009 (NFS, 2009), the estimated dose to an MEI located 300 m north-northeast of the site was 0.0018 mrem; the maximum organ doses were 0.0068 mrem (spleen) and 0.0022 mrem (red bone marrow) (doses are expressed as CEDE). Doses were calculated using reported stack effluents and a 5-year average wind rose (Class D). For 2004-2007, doses (again expressed as CEDE) to the MEI averaged only 0.007 mrem/yr (NFS 2009 license renewal12).

Airborne effluents from NFS have been decreasing since 1989. In 1999, the maximum CEDE was 2.6 mrem/yr (2.4 from air, 0.5 liquid) and the maximum lung dose was 21 mrem (NFS, 1999). External (direct) exposure was generally negligible (inhalation dose × 10–6). Internal dose was mainly from technetium-99, thorium, and uranium. There were no reported drinking water impacts for that year (NFS, 1999).

The Westinghouse fuel production facility similarly reported that the

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

critical dose pathway is inhalation (lung dose with an annual TEDE dose in 2002 of image0.4 mrem to an exposed individual living at the site boundary). The dose from liquid effluents was estimated as image0.0003 mrem/yr (Westinghouse, 2002).

3.4 APPROACHES FOR ESTIMATING DOSES FOR AN EPIDEMIOLOGIC STUDY

As discussed in Section 3.1, the use of an MEI dose is not appropriate for epidemiologic studies: MEI doses are calculated by facility licensees to demonstrate compliance with applicable regulations. It provides an estimate of dose at a single point and does not provide any information on the variation of dose as a function of distance and direction from a facility. Further, MEI doses are larger than would likely be received by any actual individual living around a nuclear facility as a result of radioactive effluent releases. More realistic estimates of individual dose as a function of distance and direction from the facility are needed to support an epidemiologic study.

Also as noted in Section 3.1, computer models have been developed to estimate absorbed doses in persons exposed to radiation through environmental pathways (see NCRP, 2009b). Such models could be used to estimate doses to individuals living near nuclear facilities to support an epidemiologic study. An existing computer model could be modified for this purpose, or a new model could be developed. Regardless of the approach used, it is essential that the computer model reflect modern practices for dose reconstruction.

Guidance provided in USNRC Regulatory Guides 1.109, 1.111, and 1.113 (USNRC, 1977a, b, c) is used by nuclear plant licensees to estimate equivalent doses to the MEI. This guidance can also be used to estimate equivalent doses to representative individuals in the vicinity of the nuclear plant. For example, a computer program was developed by PNL to estimate doses received via airborne and waterborne pathways by representative individuals living in the vicinity of operating nuclear plants from 1975 through 1988 (Baker, 1996). It is possible that this program (or similar more recent programs developed by the USNRC or other organizations) could be modified to obtain dose estimates to support the epidemiologic studies that are recommended in this report (see Chapter 4). The remainder of this section describes the modifications that would need to be made to make the PNL computer model usable for developing dose estimates to support an epidemiologic study.

It is not the intention of the committee to endorse the PNL model or to recommend its use. It is only for practical reasons that the PNL model and, by extension, the USNRC Regulatory Guides 1.109, 1.111, and 1.113 are used as a basis for the presentation of recommended modifications and

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

improvements. Namely, the PNL model was developed to use the effluent data that are reported to the USNRC by facility licensees. As noted in Chapter 2, these data represent summed quantities typically over periods of weeks to months.

The PNL model was used to estimate equivalent doses for representative individuals of population groups living in 160 segments around nuclear plants defined by 22.5-degree radial slices of the 16 compass points (i.e., N, NNE, NE, ENE, E, ESE, SE, SSE, S, SSW, SW, WSW, W, WNW, NW, NNW) and 10 concentric intervals from 2 to 80 km from the facility boundary (Table 3.8). The population was divided into four age groups: infants (image1 year), children (1-10 years), teenagers (11-17 years), and adults (> 17 years). Doses to selected organs (Table 3.9) were calculated for both airborne and waterborne pathways (Table 3.10) for 83 radionuclides (Table 3.11). The dose to a representative individual of a given age is assumed to be the same in any location within a given segment, except when the dose to the MEI was calculated.

TABLE 3.8 Concentric Intervals and Midpoints Used for Dose Calculations in the PNL Model


Distance Interval from the Plant Boundary (km) Midpoint of Interval (km)

2-3 2.5
3-4 3.5
4-6 5
6-9 7.5
9-14 11.5
14-20 17
20-30 25
30-40 35
40-60 50
60-80 70

SOURCE: Baker (1996).

TABLE 3.9 Doses to Organs Estimated by the PNL Model


Airborne Pathways Waterborne Pathways

Total body Total body
Thyroid Thyroid
Bone Bone
Gastrointestinal tract Gastrointestinal tract
Liver Liver
Lung

SOURCE: Baker (1996).
Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

TABLE 3.10 Environmental Pathways Considered in the PNL Model


Airborne Pathways Waterborne Pathways

Air submersion Ingestion of drinking water
Ground irradiation Ingestion of fish and invertebrates
Inhalation Shoreline irradiation (for MEI)
Ingestion of foodstuffs and animal products Ingestion of irrigated food products (for MEI)
Gamma and beta air dose (for MEI at site boundary)

SOURCE: Baker (1996).

TABLE 3.11 Radionuclides Considered in the PNL Model

Noble gases: 41Ar, 83mKr, 85mKr, 85Kr, 87Kr, 88K r, a89Kr, 131mXe, 133mXe, 133Xe, 135mXe, 135Xe, 137Xe, 138Xea
Radioiodines and precursors: 132Te , a133mTe , a131I,a132I, 133I,a134I, 135Ia
Other radionuclides: 3H, 10Be, 14C, 13N, 18F , 22Na, 46Sc, 51Cr, 54Mn, 56Mn, 55Fe, 59Fe, 57Co, 58Co, 60Co, 57Ni, 63Ni, 65Ni, 64Cu, 65Zn, 69mZn,a76As, 82Br, 88Rb, 89Rb,a89Sr, 90Sr, a91Sr, 92Sr, 90Y, 91mY, 95Zr, a97Zr, a95Nb, 97Nb, 99Mo,a 99mTc, 103Ru,a106Ru,a110mAg,a 115mCd, 115Cd, 125Sn,a124Sb, 125Sb,a 134Cs, 136Cs, 137Cs,a 138Cs, 139Cs,a 139Ba, 140Ba,a 140La, 141La, 141Ce, 144Ce,a 152Eu, 154Eu, 187W, 232Th,a 239Np

aThe dose calculation includes the contributions from the decay products.

SOURCE: Baker (1996).

The PNL model was developed about 30 years ago, and some of the approaches used to obtain dose estimates are outdated. Consequently, the model would need to be modified to make it useable in a modern epidemiologic study. Needed modifications are discussed below, using as a framework a general form of the calculation of the radiation dose, D, resulting from releases of radioactive materials into the environment (Till and Grogan, 2008):

D = (A × T × E × K)u, v(2)

in which

A = radionuclide activity released into the environment;

T = environmental transport, resulting in estimates of radionuclide concentrations in air, soil, water, and foodstuffs;

E = exposure factors, resulting in estimates of doses in air and of radionuclide intakes of radionuclide-contaminated air, water, and foodstuffs;

K = conversion to organ or tissue dose;

u, v = uncertainty and validation, which should be taken into account throughout the dose estimation process.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

The following subsections describe the needed modifications for each of these factors.

3.4.1 Dose (D)

Several factors are required to estimate the dose term (D) in the equation. These include the radionuclides that are released from the facility, their environmental pathways, the locations and ages of representative individuals who are exposed to these radionuclides, the specific organs exposed, and the type of dose that is estimated. These factors are described briefly in the following subsections.

3.4.1.1 Radionuclides

All of the radionuclides present in detectable quantities in the effluents released from nuclear plants appear to have been considered in the PNL model (Table 3.10). However, radionuclides released from fuel-cycle facilities, namely uranium-238 and its decay products, will also need to be included in the model if these facilities are considered in the epidemiologic study.

3.4.1.2 Environmental Pathways

The environmental pathways used in the PNL model (see Table 3.9) are adequate to estimate doses for an epidemiologic study. However, the underlying computer code would need to be modified to include doses received from direct radiation from onsite sources, from external irradiation from the shoreline of a contaminated water body, and from internal irradiation due to the consumption of irrigated food products where these doses comprise greater than 1 percent of the total dose.

3.4.1.3 Location of Representative Individuals

As noted previously, the PNL model estimates doses to representative individuals in each of 160 segments surrounding a nuclear plant. However, the spatial area of interest for an epidemiologic study (see Chapter 4) is the census tract, not the PNL segments. The PNL model could be modified to estimate doses in census tracts around nuclear facilities. For this purpose, a simplifying assumption could be made that the dose calculated at the centroid 13of the census tract is representative of the dose received at any

13 The centroid location could be determined geographically or based on population distribution.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

location in that census tract. Alternative approaches employing modern Geographic Information System (GIS) methods could also be employed to generate predicted doses on a GIS grid.

3.1.4.4 Ages of Representative Individuals

As noted previously, four age groups were considered in the PNL model but no gender distinctions were made. With respect to the estimation of doses from external irradiation, data in ICRP Publication 74 (ICRP, 1997) indicate that differences of about 30 percent between external doses to infants and adults are plausible; such differences would need to be taken into account in an epidemiologic study. With respect to the estimation of doses from internal irradiation, age and gender groups considered by the ICRP (1990) could be used: newborn (image1 year), infants (1-2 years), young children (3-7 years), older children (8-12 years), teenagers (13-17 years), adult males, and adult females.

3.1.4.5 Organs

Because one of the committee’s recommended epidemiologic study designs involves assessment of risks for all cancers (see Chapter 4), doses from internal radiation to all organs and tissues considered by the ICRP to be radiosensitive (i.e., adrenals, bladder, bone marrow, bone surface, brain, breast, esophagus, stomach, small intestine, colon, extrathoratic tissue, gall bladder, gonads, heart, kidneys, liver, lung, lymphatic nodes, muscle, oral mucosa, pancreas, prostate [males only], salivary glands, spleen, skin, thymus, thyroid, uterus/cervix [females only]) will need to be considered (ICRP, 2007b). With regard to the doses from external irradiation, the simplifying assumption could be made that all soft tissues of the body receive the same dose and that there is no age or gender dependency. However, special consideration would be warranted for red bone marrow or bone surfaces in case they are tissues of interest in an epidemiologic study.

3.1.4.6 Type of Dose

The PNL model estimates the committed equivalent dose per year of effluent release for representative individuals resulting from internal radiation. The dose of interest in epidemiologic studies is the annual absorbed dose by year of effluent release. This difference may pose a problem for long-lived alpha emitters that are released from fuel-cycle facilities because (1) the committed equivalent dose will have to be broken down into its yearly components, and (2) the dose from alpha particles will have to be separated from the dose from photons and electrons. Data files published

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

by the USEPA (USEPA, 2002) may be used to satisfy both purposes. In case of external irradiation involving gamma radiation, such problems do not exist. This problem can be avoided by modifying the model to estimate annual absorbed dose.

Doses for representative individuals are calculated using the simplifying assumptions that those individuals resided at the same place during the entire period of exposure. However, if a case-control study is carried out, doses will need to be calculated for specific individuals. It would then be important to gather information on their residential histories, at the census-tract level, of those individuals during the entire period of exposure.

3.4.2 Activities Released (A)

As noted in Chapter 2, nuclear plants and fuel-cycle facilities release different types of radionuclides and have different effluent release reporting requirements.

3.4.2.1 Nuclear Plants

As indicated in Chapter 2, the effluent releases of specific radionuclides are available on a monthly, quarterly, semiannual, or annual basis for any year since 1975. It is important to note for almost all reactors the released activities of carbon-14 are not included in the reports. Prior to 1975, when the released activities were much higher than in recent years, the information on released activities is more limited: it usually consists of total activities grouped into categories; the categories for airborne effluent releases are (1) noble gases and (2) iodine-131 and particulates with half-lives longer than 8 days. For waterborne effluent releases, the categories are (1) tritium (hydrogen-3) and (2) mixed fission and activation products. Information on the activities released for specific radionuclides appears to be only available for some reactors and some years of operation (see, for example, Logsdon and Robinson, 1971; BNL, 1979).

For the purposes of an epidemiologic study, it is essential to use reliable data for specific radionuclides. For most reactors and years before 2010, the airborne releases of carbon-14 in the form of CO2 will have to be estimated, for example on the basis of the thermal power generated or according to methods developed by EPRI (2010) or the USNRC (1979). Because there is no easy way to trap CO2, it is presumed that practically all of the carbon-14 activity that is produced as CO2 is released into the atmosphere. For years prior to 1975, simplifying assumptions might have to be made to reconstruct the released activities of some radionuclides14;

14 For example, the activities of individual radionuclides might have to be estimated using radionuclide distributions and group activities.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

the uncertainties attached to the estimates of reconstructed activities for specific radionuclides, which may be very large, will have to be evaluated.

Another consideration is the time period over which the activities are summed (i.e., monthly, quarterly, semiannually, annually, or by batch) for the purposes of dose estimation. The decision over which time period to select may vary from site to site and from year to year according to the availability of other data that are needed for dose estimation, such as meteorological data and river flow data. In any case, the doses of interest for the epidemiologic study are annual doses. Consequently, any doses estimated for any fraction of the year will have to be summed over the entire year.

It is worth noting that the doses from direct radiation due to nitrogen-16 contained in BWRs and radioactive materials stored onsite do not depend on the activities released, but rather on the shielding characteristics of the reactor and its procedures for storing waste materials. The corresponding doses will have to be based on site-specific measurements or on site-specific calculations.

3.4.2.2 Fuel-Cycle Facilities

At this time, the information that will be available for the entire period of operation of any fuel-cycle facility is unclear (see Chapter 2), as it seems that at least part of this information will have to be requested from the plant licensees. Annual releases of specific radionuclides would be needed to calculate doses using the PNL model.

3.4.3 Environmental Transport (T)

Environmental transport parameters link the radionuclide activities released to the concentrations of those radionuclides in environmental media (air, soil, water, sediments, and food products) at any time and location in the vicinity of a nuclear facility. A list of the main environmental transport parameters is provided in Table 3.12. Transport of airborne and waterborne releases are described in the following subsections.

3.4.3.1 Airborne Effluent Releases

The most important environmental parameter for airborne effluent releases is the atmospheric dilution factor, which is the quotient of the radionuclide concentration at the location of interest (expressed, for example, in Ci m–3) and the release rate of that radionuclide (expressed, for example, in Ci s–1). In the PNL model, atmospheric dilution factors are calculated as averages over 160 segments and also for specific locations near the plant site (site boundary, closest residence, closest garden, and closest pasture).

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
×

TABLE 3.12 Main Parameters Used to Estimate Dose per Unit Activity Released.


Pathway of Exposure Environmental Transport (T) Exposure Factors (E) Conversion to Organ or Tissue Dose (K)

Airborne Effluent Releases
     Air submersion Atmospheric dilution factor Indoor shielding and occupancy factors Dose coefficient (FGR 12)
     Ground irradiation Atmospheric dilution factor; dry deposition velocity Indoor shielding and occupancy factors Dose coefficient (FGR 12)
     Direct radiation Transport model Indoor shielding and occupancy factors Dose coefficient (ICRP 74)
     Inhalation Atmospheric dilution factor Indoor shielding and occupancy factors; breathing rates Dose coefficients (ICRP 71)
     Ingestion Atmospheric dilution factor; dry deposition velocity; transfer coefficients Consumption rates; culinary factors; holdup times Dose coefficients (ICRP 56, 67, 69)
Waterborne Effluent Releases
     Ingestion (water) Aquatic dilution factor Consumption rate; water treatment Dose coefficients (ICRP 56, 67, 69)
     Ingestion (fish and invertebrates) Aquatic dilution factor; transfer coefficients Consumption rates; culinary factors; holdup times Dose coefficients (ICRP 56, 67, 69)
     Ingestion (irrigated products) Aquatic dilution factor Consumption rates; culinary factors; holdup times Dose coefficients (ICRP 56, 67, 69)
     Shoreline irradiation Transport model Occupancy factor Dose coefficient (FGR12)

NOTE: FGR, Federal Guidance Report; ICRP, International Commission on Radiological Protection.

Several sets of atmospheric dilution factors are calculated according to the height of effluent release: ground, elevated, or mixed mode. Several assumptions are made about depletion15 and radioactive decay.

15 Depletion reflects the loss of activity in the radioactive cloud along its transport downwind as a result of radioactive decay and deposition on the ground via dry (sedimentation or impaction) or wet (rain or snow) processes.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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The values for atmospheric dilution factors are derived from sets of meteorological data that are recorded by the licensee on an hourly basis: wind speed, wind direction, and atmospheric stability class. These meteorological data are averaged over a specific year (or over a period of time greater than one year) that differed from one plant to another to obtain annual joint frequency distributions.

For the purposes of the epidemiologic study, it seems sufficient for recent years of effluent release to use the annual average atmospheric dilution factors calculated for the appropriate release height(s) using the correction for depletion and decay according to the physical half-life radionuclide that is considered. For early years (prior to 1975), calculation of the atmospheric dilution factors over the year of release that is considered or averaged on a quarterly basis or for the time of the batch releases during that same year could be considered if the appropriate meteorological data are available. In case the meteorological data are not available for the year or time period of interest, data averaged over 5-year time periods representative of the time period or year of interest could be used.

As shown in Table 3.12, the atmospheric dilution factor is the only environmental transport parameter that is needed to calculate the doses resulting from air submersion and inhalation. With respect to the doses from ground irradiation and ingestion, the radionuclide activities deposited per unit area of ground (expressed, for example, in Ci m–2) are needed. In the PNL model, activities on the ground are also derived from the annual joint frequency distributions, supplemented with values of dry deposition velocity (a quantity that relates the activity deposited on the ground to the ground-level air concentration, in the absence of precipitation). For the purposes of the epidemiologic study, the same procedure could be used. It is recognized that the influence of the precipitation events, which are more effective than dry processes in scavenging the radioactive materials from the atmosphere, would not be taken into account. This is deemed to be a reasonable simplification because deposition on the ground does not occur for noble gases and occurs by different processes for tritium and carbon-14, which are the most important contributors to the dose from airborne releases. Finally, the deposition on the ground is partitioned between the activity that is first retained by vegetation and the activity that falls directly on the soil.

With respect to ingestion of food products, the activity deposited on the ground must be related to the radionuclide concentrations in agricultural products (mainly milk, leafy vegetables, and meat). This is done by means of transfer coefficients. Those provided in Tables E.1 and E.2 of Regulatory Guide 1.109 (USNRC, 1977a) should not be adopted blindly: In the framework of an epidemiologic study, it would be important to carry out a thorough literature search, especially for tritium and carbon-14, which

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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seem to be the most important radionuclides with respect to intakes by ingestion, to determine which coefficients to use.

3.4.3.2 Waterborne Effluent Releases

Just as in the case of the atmospheric effluent releases, the most important parameter in waterborne releases is the aquatic dilution factor, which is the quotient of the radionuclide concentration at the location of interest (expressed, for example, in Ci m–3) and the release rate of that radionu-clide (expressed, for example, in Ci s–1). The locations of interest are those where water is taken for drinking or irrigation purposes (for freshwater releases) and where fish and invertebrates are harvested (for saltwater as well as for freshwater releases). For releases into rivers, the aquatic dilution factor can be reasonably assumed to correspond to homogeneous mixing of the released activity into the entire flow of the river. For other types of releases (into lakes, estuaries, oceans, etc.), the aquatic dilution factors are site specific.

In the PNL model, the annual average values of the aquatic dilution factors are, whenever possible, taken from the environmental information provided by the licensees; when no information is available, the PNL model provides default values. For the purposes of the epidemiologic study, it also seems sufficient to use annual averages of the aquatic dilution factors. Whenever possible, site-specific values should be derived from a thorough analysis of the relevant documentation.

With respect to ingestion of fish and invertebrates, the radionuclide concentrations in those foodstuffs are derived from the radionuclide concentrations in water using transfer coefficients, for example expressed in Ci kg–1 or Ci m–3. Element-specific recommended values of such transfer coefficients, in the absence of site-specific data, are listed in Table A.1 of Regulatory Guide 1.109 (USNRC, 1977a). If site-specific data are not available, more up-to-date transfer coefficients may be available from other sources.

3.4.4 Exposure Factors (E)

For each pathway, exposure factors, representing the usage that humans make of the environment and of its products, have to be taken into consideration. In the PNL model, site-dependent parameter values were taken from plant-specific environmental information whenever possible. However, site-dependent values were usually not available; in that case, the generic values recommended in Regulatory Guide 1.109 (USNRC, 1977a) and presented in Table 3.13 were used.

The values of the exposure factors presented in Table 3.13 for inhalation and external irradiation appear to be reasonable for use in an

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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TABLE 3.13 Generic Values of Exposure Factors Used in the PNL Model for Average Members of the Population


Pathway Infant Child Teenager Adult

Ingestion: milk (L yr-1) 170 170 200 110
Ingestion: meat and poultry (kg yr-1) 0 37 59 95
Ingestion: fruits, vegetables, and grains (kg yr-1) 0 200 240 190
Ingestion: water (L yr-1) 170 260 260 370
Ingestion: fish (kg yr-1) 0 2.2 5.2 6.9
Ingestion: invertebrates (kg yr-1) 0 0.33 0.75 1.0
Inhalation: breathing rate (m-3 yr-1) 1400 3700 8000 8000
External irradiation: shielding and occupancy factor 0.5 0.5 0.5 0.5

SOURCE: Based on Table A-1 in NUREG/CR 2850, vol. 1 (1982).

epidemiologic study. For ingestion, however, two important considerations are not taken into account: (1) the fact that water treatment and culinary processes may result in a decrease in radionuclide concentrations in the consumed water and food products, and (2) the dilution of contaminated water and food products due to consumption of water and food products from noncontaminated sources. These factors would need to be taken into account in the framework of an epidemiologic study.

3.4.5 Conversion to Organ or Tissue Dose (K)

The conversion factors used to calculate doses from the activity intakes of water and food products (in the case of internal irradiation), and, in the case of external irradiation, from the ground-level air concentrations weighted according to shielding and indoor occupancy (for air submersion), and from the radionuclide concentrations in soil and sediments (for ground irradiation and shoreline irradiation, respectively) are discussed in Appendix I. Generally speaking, the factors related to external irradiation appear to be adequate for use in an epidemiologic study, but those related to internal irradiation will have to be updated with data included in the publications of the ICRP-56 series (ICRP, 1990, 1992, 1995a, b). These ICRP data are in terms of committed equivalent doses per unit intake. Additionally, because it will be important to calculate annual absorbed doses for high-LET and low-LET radiations separately, it will be necessary, for radionuclides with long biological half-lives of residence in the body (e.g., strontium-90) and for all alpha emitters, to use data files published by the USEPA (USEPA, 2002) that provide the required information. For all other radionuclides, the committed equivalent doses per unit intake are numerically

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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equal to the annual absorbed doses per unit intake, so that the data provided in the publications of the ICRP-56 series can be used without modification.

3.5 OTHER RISK FACTORS

Individuals living near nuclear facilities may be exposed to radiation from other sources besides facility effluent releases. The most significant sources of these other exposures are from natural background radiation, radiation from medical diagnostic procedures, and cosmic radiation from air travel. For the purposes of dose reconstruction, all radiation is equal: That is, a cell, tissue, or organ cannot distinguish between radiation received from USNRC-licensed facilities and radiation received from these other sources. In fact, these other sources of radiation exposure may result in doses that are much larger than those from facility effluent releases. If doses from these other sources are differentially distributed in individuals living near a nuclear facility (e.g., by distance or direction from a facility), they could confound the results of an epidemiologic study (see Chapter 4). Even if these doses are not differentially distributed, they would still produce “noise” that could swamp the “signal” resulting from exposures to facility effluent releases. In either case, these other sources of exposure are risk factors that need to be considered in dose assessment studies.

3.5.1 Natural Background Radiation

As noted in Section 3.2, reported annual whole-body doses from nuclear facilities were generally at most only 10-20 mrem/yr to the MEI (e.g., Table 3.2), even in early years of facility operations when effluent levels were much higher than at present. Reported average doses to populations living within a few miles of a plant were generally much less than 1 mrem/ yr. These doses are much lower than annual whole-body absorbed doses received from natural background radiation.

The levels of terrestrial gamma radiation from naturally occurring radioactivity in soil and building materials and from cosmic rays vary widely across the United States (NCRP, 2009a). For example, free-in-air terrestrial gamma radiation levels measured at 210 sites in the United States averaged 61 mrad/yr with a standard deviation of 23 mrad/yr (Eisenbud and Gesell, 1997).

Cosmic radiation adds to natural background levels. Cosmic-ray levels vary with altitude from about 30 mrad/yr at sea level to over 50 mrad/ yr at high altitudes (Lowder and Beck, 1966; NCRP, 2009a). Thus, direct external radiation doses to persons living near nuclear plants due to facility effluents were much less than the doses they received from ambient natural

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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background at most sites, even in the 1970s and 1980s. They were also generally much less than the spatial and temporal variations in natural background radiation from site to site.

The natural terrestrial background radiation level at any site in any annual quarter can vary by several mrad due to variations in rainfall (soil moisture), snow cover, and radon levels. Figure 3.7 illustrates daily variations in exposure rate measured at a site in New Jersey in 1979.

The natural background doses cited above are free in air (that is, uncorrected for shielding by housing and indoor radiation sources). The exact dose to any individual from facility releases would depend very much on their exact location when the releases occurred, type of housing (shielding), the fraction of time an individual spent in housing or away from the facility vicinity, and other factors. The doses cited above also do not include internal exposure from naturally occurring radionuclides in the body or exposure to indoor radon.

Background doses from terrestrial and cosmic-ray free-in-air external exposure have been estimated only for some selected facilities using those facilities’ reported TLD monitoring data. The approximate annual terrestrial background exposures16 are shown in Figure 3.8. These annual background doses often vary by more than a factor of 3, and they are one or more orders of magnitude higher than the estimated doses to the MEI discussed elsewhere in this chapter (e.g., Table 3.2).

The spatial variations in background can be significant even over relatively small distances. Figure 3.9 shows the spatial variation, based on annual TLD readings, around the Millstone plant in 2009 when external radiation exposures due to effluents from the facility in 1979 were essentially zero. Annual background radiation levels varied by over a factor of 2 and were higher west of the plant than north of the plant. Variations over shorter intervals were likely even greater.

Because the ambient background doses are so much higher than expected doses from facility effluent releases and vary both with direction and distance, the epidemiologic study will need to consider variations in background radiation not only from facility to facility, but also around each facility. By evaluating the reported quarterly TLD monitoring data from each facility for recent years (when facility contributions to dose were very low), reasonable estimates of average annual background doses as a function of distance and direction can be made for use in the epidemiologic study.

16 Based on the facility TLD monitors (biased low due to partial shielding because TLDs are generally attached to telephone poles, trees, or buildings). Note that these “background” exposures do not include exposures from internal emitters or indoor radon.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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image

FIGURE 3.7 Daily variations in background radiation for a site in New Jersey. SOURCE: Beck and Miller (1982).

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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image

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.

image

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, Inc. (2009).

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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3.5.2 Other Sources of Radiation

Individuals living near nuclear facilities receive radiation from a number 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 exposed 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 epidemiologic 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 protons

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.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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(i.e., hydrogen nuclei) and alpha particles (i.e., helium nuclei). These particles interact with air molecules in the atmosphere and generate additional 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 studies 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 uncertainties 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 nuclear 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 radioactivity (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 vicinity 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.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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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 concentrations 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 effluent 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 effluent releases. The airborne effluent release uncertainties are a function of how representative the weekly grab samples22 were with respect to the actual 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 exposure pathway models relating environmental concentrations to doses can be high. Atmospheric dispersion estimates can also be very uncertain, particularly 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 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 annual averages and result in significant errors if the latter are used to estimate doses for batch effluent releases into the atmosphere.

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 Appendix H.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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TABLE 3.14 Uncertainties in Gaussian Plume Models


Conditions Range, Predicted over Observed
Air Concentration (P/O)

Highly instrumented site; ground-level, centerline; within 10 km of a continuous point source 0.65 to 1.35
Specific time and location, flat terrain, steady meteorology, within 10 km of release point 0.1 to 10
Annual average, specific location, flat terrain, within 10 km of release point 0.5 to 2
Annual average, specific location, flat terrain, 10 m to 150 km downwind 0.25 top 4
Complex terrain or meteorology, episodic releases 0.01 to 100
Episodic, surface-level releases, wind speeds less than 2 m s -1 1 to 100

SOURCE: Miller (1995).

Liquid diffusion models for effluent releases into estuaries, lakes, and oceans, as well as spills into surface and ground water, are very crude. Additionally, 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 reconstruction. However, a quantitative or at least semiquantitative uncertainty

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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analysis could be performed, at least for a few facilities and years of operation, 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 uncertainty 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 As discussed 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 uncertainty) 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).

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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  1. 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 levels 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 Section 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 pathways. 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 resulting 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.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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  1. 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:
  • Using the model to obtain dose estimates as a function of distance (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).
  • Developing a methodology to account for natural background radiation and, to the extent feasible, other sources of radiation in the dose estimates.
  • Undertaking an uncertainty analysis as described in Section 3.6.

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 Discharges, 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, National Toxicology Program. Report on Carcinogens, 12th Edition. Available at http://ntp.niehs.nih.gov/ntp/roc/twelfth/roc12.pdf.

Suggested Citation:"3 Radiation Dose Assessment." National Research Council. 2012. Analysis of Cancer Risks in Populations Near Nuclear Facilities: Phase 1. Washington, DC: The National Academies Press. doi: 10.17226/13388.
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In the late 1980s, the National Cancer Institute initiated an investigation of cancer risks in populations near 52 commercial nuclear power plants and 10 Department of Energy nuclear facilities (including research and nuclear weapons production facilities and one reprocessing plant) in the United States. The results of the NCI investigation were used a primary resource for communicating with the public about the cancer risks near the nuclear facilities. However, this study is now over 20 years old. The U.S. Nuclear Regulatory Commission requested that the National Academy of Sciences provide an updated assessment of cancer risks in populations near USNRC-licensed nuclear facilities that utilize or process uranium for the production of electricity.

Analysis of Cancer Risks in Populations near Nuclear Facilities: Phase 1 focuses on identifying scientifically sound approaches for carrying out an assessment of cancer risks associated with living near a nuclear facility, judgments about the strengths and weaknesses of various statistical power, ability to assess potential confounding factors, possible biases, and required effort. The results from this Phase 1 study will be used to inform the design of cancer risk assessment, which will be carried out in Phase 2. This report is beneficial for the general public, communities near nuclear facilities, stakeholders, healthcare providers, policy makers, state and local officials, community leaders, and the media.

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