Return to Index
The potential radiological dose to the public from Idaho National Laboratory (INL) Site operations was evaluated to determine compliance with pertinent regulations and limits. Two different computer models were used to estimate doses: Clean Air Act Assessment Program 1988 (CAP-88) and the mesoscale diffusion (MDIFF) air dispersion model. CAP-88 is required by the U.S. Environmental Protection Agency to demonstrate compliance with the Clean Air Act. The National Oceanic and Atmospheric Administration Air Resources Laboratory–Field Research Division developed MDIFF to evaluate dispersion of pollutants in arid environments such as those found at the INL Site. The maximum calculated dose to an individual by either of the methods was well below the applicable radiation protection standard of 10 mrem/year. The dose to the maximally exposed individual, as determined by the CAP-88 program, was 0.077 mrem (0.77 μSv). The dose calculated using the MDIFF values was 0.041 mrem (0.41 μSv). The maximum potential population dose to the approximately 281,495 people residing within a 80‑km (50-mi) radius of any INL Site facility was 0.565 person-rem (5.7 x 10-3 person-Sv), well below that expected from exposure to background radiation.
Using the maximum radionuclide concentrations in collected waterfowl, game animals, and doves, a maximum potential dose from ingestion was calculated. The maximum potential dose was estimated to be 0.19 mrem (1.9 µSv) for waterfowl and 0.005 mrem (0.05 µSv) for game animals. No potential dose would result from consuming doves collected in 2005.
The potential dose to aquatic and terrestrial biota from contaminated soil and water was also evaluated, using a graded approach. Based on this approach, there is no evidence that INL Site related radiological contamination is having an adverse impact on populations of plants and/or animals.
It is the policy of the U.S. Department of Energy (DOE) “To implement sound stewardship practices that are protective of the air, water, land, and other natural and cultural resources impacted by DOE operations and by which DOE cost-effectively meets or exceeds compliance with applicable environmental; public health; and resource protection laws, regulations, and DOE requirements” (DOE 2003). DOE Order 5400.5 further states, “It is also a DOE objective that potential exposures to members of the public be as far below the limits as is reasonably achievable...” (DOE 1993). This chapter describes the dose to members of the public and to the environment based on the 2005 radionuclide concentrations from operations at the Idaho National Laboratory (INL) Site.
Back to top
Individual radiological impacts to the public surrounding the INL Site remain too small to be measured by available monitoring techniques. To show compliance with federal regulations established to ensure public safety, the dose from INL Site operations was calculated using the reported amounts of radionuclides released during the year from INL Site facilities (see Chapter 4) and appropriate air dispersion computer codes. During 2005, this was accomplished for the radionuclides summarized in Table 4-2 (Sections A, B, C, and D).
The following estimates were calculated:
In this chapter, the term dose refers to effective dose equivalent unless another term is specifically stated. Dose was calculated by summing the effective dose equivalents from immersion, inhalation, ingestion, and deposition. Effective dose equivalent includes doses received from both external and internal sources and represents the same risk as if an individual’s body were uniformly irradiated. U.S. Environmental Protection Agency (EPA) dose conversion factors and a 50-year integration period were used in calculations in combination with the MDIFF air dispersion model for internally deposited radionuclides (Eckerman et al. 1988) and for radionuclides deposited on the ground surface (Eckerman and Ryman 1993). The CAP-88 computer code uses dose and risk tables developed by the EPA. No allowance is made in the dose calculations using MDIFF for shielding by housing materials, which is estimated to reduce the dose by about 30 percent, or less than year-round occupancy time in the community. The CAP‑88 computer code does not include shielding by housing materials, but it does include a factor to allow for shielding by surface soil contours from radioactivity on the ground surface.
Of the potential exposure pathways by which radioactive materials from INL Site operations could be transported offsite (see Figure 3-1), atmospheric transport is the principal potential pathway for exposure to the surrounding population. This is because winds can carry airborne radioactive material rapidly and some distance from its source. The water pathways are not considered major contributors to dose because no surface water flows off the INL Site and no radionuclides from the INL Site have been found in drinking water wells offsite. Because of these factors, the MEI dose is determined through the use of computer codes of atmospheric dispersion of airborne materials.
Back to top
The NESHAP, as outlined in Title 40, Code of Federal Regulations (CFR), Part 61
(40 CFR Part 61), Subpart H, requires the demonstration that radionuclides other
than radon released to air from any DOE nuclear facility do not result in a dose
to the public of greater than 10 mrem/year (EPA 2006). This includes releases
from stacks and diffuse sources. The EPA requires the use of an approved
computer code to demonstrate compliance with 40 CFR Part 61. The INL Site uses
the code CAP‑88 as recommended in 40 CFR 61 to demonstrate NESHAP compliance.
The National Oceanic and Atmospheric Administration Air Resources Laboratory–Field Research Division (NOAA ARL-FRD) developed a mesoscale air dispersion model called MDIFF (formerly known as MESODIF) (Sagendorf et al. 2001) around 1970. The MDIFF diffusion curves were developed by the NOAA ARL-FRD from tests in arid environments (e.g., the INL Site and the Hanford Site in eastern Washington). The MDIFF curves are more appropriate for estimating dose to the public caused by INL Site emissions than those used by the CAP-88 code. The MDIFF code is a dispersion model only and does not account for plume depletion and radioactive decay.
The MDIFF model is used to calculate total integrated concentrations (TICs) that are then used to calculate the dose to members of the public residing near the INL Site. In previous years, doses calculated from the MDIFF TICs have been somewhat higher than doses calculated using CAP-88. Differences between the two computer codes were discussed in detail in the 1986 annual report (Hoff et al. 1987). The primary difference is the atmospheric dispersion portion of the codes. CAP-88 makes its calculations based on the joint frequency of wind conditions from a single wind station located near the source in a straight line from that source and ignores recirculation. MDIFF calculates the trajectories of a puff using wind information from 36 towers in the Upper Snake River Plain. This allows for more accurate and site-specific modeling of the movement of a release using prevailing wind conditions between time of the release and the time that the plume leaves the INL Site boundary. For this reason, the two computer codes may not agree on the location of the MEI or the magnitude of the maximum dose.
The offsite concentrations calculated using both computer codes were compared to actual monitoring results using the radionuclide antimony-125 at offsite locations in 1986, 1987, and 1988 (Hoff et al. 1987, Chew and Mitchell 1988, Hoff et al. 1989). Concentrations calculated for several locations using the MDIFF TICs showed good agreement (within a factor of 2) with concentrations from actual measurements, with the model calculations generally predicting concentrations higher than those measured. The original computer code (MESODIF) was extensively studied and validated, and compared to other models in the mid-1980s (Lewellen, et al. 1985, Start et al. 1985, Sagendorf and Fairobent 1986).
The dose from INL Site airborne releases of radionuclides calculated to demonstrate compliance with NESHAP are published in the National Emissions Standards for Hazardous Air Pollutants-Calendar Year 2005 INL Report for Radionuclides (DOE-ID 2006). For these calculations, 63 potential maximum locations were evaluated. The CAP-88 computer code predicted the highest dose to be at Frenchman’s Cabin, located at the southern boundary of the INL Site. This location is only inhabited during portions of the year, but it must be considered as a potential MEI location according to the NESHAP. At Frenchman’s Cabin, an effective dose equivalent of 0.077 mrem (0.77 μSv) was calculated. The dose of 0.077 mrem (0.77 μSv) is well below the whole body dose limit of 10 mrem (100 μSv) for airborne releases of radionuclides established by 40 CFR 61.
Using data gathered continuously at 36 meteorological stations on and around the INL Site and the MDIFF model, the NOAA ARL-FRD prepares a mesoscale map (Figure 8-1) showing the calculated 2005 time integrated concentrations (TICs). These TICs are based on a unit release rate weighted by percent contribution for each of eight INL Site facilities: Central Facilities Area (CFA), Idaho Nuclear Technology and Engineering Center (INTEC), Materials and Fuels Complex (MFC), Naval Reactors Facility (NRF), Critical Infrastructure Test Range Complex (CITRC), Reactor Technology Complex (RTC), Radioactive Waste Management Complex (RWMC), and Test Area North (TAN). To create the isopleths shown in Figure 8-1, the TIC values are contoured. Average air concentrations (in curies per cubic meter [Ci/m3]) for a radionuclide released from a facility are estimated from a TIC isopleth (line of equal air concentration) in Figure 8-1. To calculate the average air concentration, the TIC is multiplied by the quantity of the radionuclide released (in curies [Ci]) during the year and divided by the number of hours in a year squared (8760 hour)2 or 7.67 x 107 hour2. This does not account for plume depletion, radioactive decay, or in-growth or decay of radioactive progeny.
The average air concentrations calculated by MDIFF were input into a Microsoft Excel spreadsheet program developed by the Environmental Surveillance, Education and Research (ESER) Program to calculate doses using methods outlined in U.S. Nuclear Regulatory Commission (NRC 1977) and dose conversion factors provided by EPA (EPA 2002). In 2000, a revision to the methods and values used for the calculation of the MEI dose using the MDIFF TIC values was undertaken. Values for the deposition and plant uptake rates of radionuclides, most noticeably radioiodines, were modified to reflect present operations and current values in use. The most notable change, mathematically, is the increase of the iodine‑129 (129I) deposition velocity from 0.01 m/second (0.03 ft/second) to 0.035 m/second (0.11 ft/second), as the emitted radioiodines went from predominantly organic in nature to elemental. These changes resulted in a mathematical increase in the amount of radionuclides deposited on the ground and available for plant uptake. This increase in deposited radionuclides leads to a corresponding net increase in the ingestion dose.
The MDIFF model predicted that the highest TIC for radionuclides in air at a location with a year-round resident during 2005 would have occurred northwest of Mud Lake. The maximum hypothetical dose was calculated for an adult resident at that location from inhalation of air, submersion in air, ingestion of radioactivity on leafy vegetables, and exposure because of deposition of radioactive particles on the ground. The calculation was based on data presented in Table 4-2 (Sections A, B, C, and D) and the grid used to produce Figure 8-1.
Using the largest calculated TIC for each facility (Table 8-1) at the location inhabited by a full-time resident, and allowing for radioactive decay and plume depletion during the transit of the radionuclides from each facility to the location of the MEI (northwest of Mud Lake), the potential annual effective dose equivalent from all radionuclides released was calculated to be 0.041 mrem (0.41 μSv) (Table 8-2). This dose is well below the whole body dose limit of 10 mrem set in the 40 CFR 61 for airborne releases of radionuclides.
For 2005, the ingestion pathway was the primary route of exposure and accounted for 72 percent of the total dose, followed by inhalation at 26 percent, and immersion at 2 percent. Deposition accounted for only 0.22 percent of the dose.
Radionuclide releases for 2005 are presented in Figure 8-2. The noble gas krypton-85 (85Kr) accounted for approximately 78 percent of the total release, followed by argon-41 (41Ar) with 9 percent, and tritium at 8 percent of the total. The noble gases xenon-133 (133Xe) and -135 (135Xe) contributed 0.4 and 0.3 percent, respectively. However, because these are noble gases they contribute very little to the cumulative dose (affecting immersion only). Other than 41Ar and tritium (3H), the radionuclides contributing to the overall dose were 0.02 percent of the total radionuclides released.
The largest contributor to the MEI dose was cesium-137 (137Cs), accounting for 37.3 percent of the total dose (Figure 8-3). This was followed by strontium-90 (90Sr) at 21.6 percent and americium-241 (241Am) at 11.6 percent. Isotopes of plutonium (plutonium-238 [238Pu], plutonium-239 [239Pu], plutonium-240 [240Pu], and plutonium-241 [241Pu]) contributed a total of 15.1 percent to the dose.
The respective contribution to the overall dose by facility is as follows: TAN (45 percent), INTEC (35 percent), RTC (16 percent), and RWMC (3 percent). NRF contributed approximately 0.13 percent of the 2005 total dose, while MFC contributed about 0.11 percent. CFA and CITRC each accounted for less than 0.01 percent of the total dose.
The calculated maximum dose resulting from INL Site operations is still a small fraction of the average dose received by individuals in southeastern Idaho from cosmic and terrestrial sources of naturally occurring radiation found in the environment. The total annual dose from all natural sources is estimated at approximately 358 mrem (Table 7-12).
Table 8-3 summarizes the calculated annual effective dose equivalents for 2005 from INL Site operations using both the CAP‑88 and MDIFF air dispersion computer codes. A comparison is shown between these doses and the EPA airborne pathway standard and the estimated dose from natural background. The reasons for the disparity in the MDIFF and CAP‑88 doses are a result of the changes made to the calculations discussed above.
As with the calculation of the maximum individual dose, the determination of the population dose also underwent changes in 2000. Using the power of a geographical information system (ArcView), annual population no longer needs to be distributed using growth estimations and a specialized computer code. In addition to this simplification, the population dose is now calculated for the population within an 80 km (50 mi) radius of any INL Site facility. This takes into account the changes in facility operations, in that the INTEC is not always the single largest contributor of radionuclides released.
An estimate was made of the collective effective dose equivalent, or population dose, from inhalation, submersion, ingestion, and deposition resulting from airborne releases of radionuclides from the INL Site. This collective dose included all members of the public within 80 km (50 mi) of an INL Site facility. The population dose was calculated in a spreadsheet program that multiplies the average TIC for the county census division (in hours squared per cubic meter) by the population in each census division within that county division and the normalized dose received at the location of the MEI (in rem per year per hour squared per meter cubed). This gives an approximation of the dose received by the entire population in a given county division (Table 8-4).
The dose received per person is obtained by dividing the collective effective dose equivalent by the population in that particular census division. This calculation overestimates dose because the model conservatively does not account for radioactive decay of the isotopes during transport over distances greater than the distance from each facility to the residence of the MEI located northwest of Mud Lake. Idaho Falls, for example, is about 50 km (31 mi) from the nearest facility (MFC) and 80 km (50 mi) from the farthest. Neither residence time nor shielding by housing was considered when calculating the MEI dose on which the collective effective dose equivalent is based. The calculation also tends to overestimate the population doses because they are extrapolated from the dose computed for the location of the potential MEI. This individual is potentially exposed through ingestion of contaminated leafy garden vegetables grown at that location.
The 2005 MDIFF TIC used for calculation of the population dose within each county division were obtained by averaging the results from appropriate census divisions contained within those county divisions. The total population dose is the sum of the population doses for the various county divisions (Table 8-4). The estimated potential population dose was 0.565 person-rem (5.7 x 10-3 person-Sv) to a population of approximately 286,144. When compared with an approximate population dose of 102,439 person-rem (1024 person-Sv) from natural background radiation, this represents an increase of only about 0.0005 percent. The largest collective doses are found in the Idaho Falls and Pocatello census divisions due to their greater populations.
Back to top
The potential dose an individual may receive from the occasional ingestion of meat from game animals continues to be investigated at the INL Site. Such studies include the potential dose to individuals who may eat (1) waterfowl that reside briefly at wastewater disposal ponds at RTC, INTEC, and MFC that are used for the disposal of low-level radioactive wastes and (2) game birds and game animals that may reside on or migrate across the INL Site.
A study was initiated in 1994 to obtain data on the potential doses from waterfowl using INL Site wastewater disposal ponds. This study focused on the two hypalon-lined evaporation ponds at RTC that replaced the percolation ponds formerly used for disposal of wastewater at that facility (Warren et al. 2001).
In the summer of 2005, three ducks were collected from the RTC wastewater ponds, three were collected from wastewater ponds at the MFC, and three were collected from an offsite location (near Firth, Idaho) as controls. The maximum potential dose from eating 225 g (8 oz) of meat from ducks collected in 2005 is presented in Table 8-5. Radionuclide concentrations used to determine these doses are reported in Table 7-4. Doses from consuming waterfowl are based on the assumption that ducks are eaten immediately after leaving the ponds.
The maximum potential dose of 0.19 mrem (1.9 µSv) from these waterfowl samples, while higher than those in the past few years, is substantially below the 0.89 mrem (8.9 µSv) committed effective dose equivalent estimated from the most contaminated ducks taken from the evaporation ponds between 1993 and 1998 (Warren et al. 2001). The ducks were not collected directly from the hypalon-lined radioactive wastewater ponds but from the adjacent sewage lagoons. However, the birds likely used the radioactive wastewater ponds during the approximate two-week period they were observed in the area.
No manmade radionuclides were found in any of the three mourning dove samples collected in 2005. Therefore, there was no potential dose from manmade radionuclides received from eating these doves.
Big Game Animals
A conservative estimate of the potential whole-body dose that could be received from an individual eating the entire muscle (27,000 g [952 oz]) and liver mass (500 g [17.6 oz]) of an antelope with the highest levels of radioactivity found in these animals was estimated at 2.7 mrem in a study on the INL Site from 1976-1986 (Markham et al. 1982). Game animals collected at the INL Site during the past few years have shown much lower concentrations of radionuclides. Only one game animal collected during 2005 had a detectable concentration of 137Cs in the muscle; none had a detectable concentration in liver tissue. Based on the concentration of 137Cs found in the muscle of this game animal, the potential dose was approximately 0.005 mrem (0.05 µSv).
The contribution of game animal consumption to the population dose has not been calculated because only a limited percentage of the population hunts game, few of the animals killed have spent time on the INL Site, and most of the animals that do migrate from the INL Site would have reduced concentrations of radionuclides in their tissues by the time they were harvested (Halford et al. 1983). The total population dose contribution from these pathways would, realistically, be less than the sum of the population doses from inhalation of air, submersion in air, ingestion of vegetables, and deposition on soil.
Back to top
The impact of environmental radioactivity at the INL on nonhuman biota was assessed using the graded approach procedure detailed in A Graded Approach for Evaluating Radiation Doses to Aquatic and Terrestrial Biota (DOE 2002) and the associated software, RESRAD-Biota (ISCORS 2004). The graded approach evaluates the impacts of a given set of radionuclides on aquatic and terrestrial ecosystems by comparing available concentration data in soils and water with biota concentration guides (BCGs). A BCG is defined as the environmental concentration of a given radionuclide in soil or water that, under the assumptions of the model, would result in a dose rate less than 1 rad/day (10 mGy/day) to aquatic animals or terrestrial plants or 0.1 rad/day (1 mGy/day) to terrestrial animals. If the sum of the measured environmental concentrations divided by the BCGs (the combined sum of fractions) is less than one, no negative impact to populations of plants or animals is expected. No doses are calculated unless the screening process indicates a more detailed analysis is necessary.
The approach is graded because it begins the evaluation using conservative default assumptions and maximum values for all currently available data. Failure at this general screening step does not necessarily imply harm to organisms. Instead, it is an indication that more realistic model assumptions may be necessary. Several specific steps for adding progressively more realistic model assumptions are recommended. After applying the recommended changes at each step, if the combined sum of fractions is still greater than one, the graded approach recommends evaluating the next step. The steps can be summarized as:
Each step of this graded approach requires appropriate justification before it can be applied. For example, before using the mean concentration, assessors must discuss why the maximum concentration is not representative of the radionuclide concentration to which most members of the plant or animal population are exposed.
Evaluations beyond the initial general screening require assessors to make decisions about assessment areas, organisms of interest, and other factors. Of particular importance for the terrestrial evaluation portion of the 2005 biota dose assessment is the division of the INL Site into evaluation areas based on potential soil contamination and habitat types (Figure 8-4). Details and justification are provided in Morris (2003).
The graded approach (DOE 2002) and RESRAD-Biota (ISCORS 2004) are designed to evaluate certain common radionuclides. Thus, this biota dose assessment evaluated potential doses from radionuclides detected in soil or water on the INL that are also included in the graded approach (Table 8-6).
For this analysis, maximum effluent data were used because actual pond water samples were not available. These data are assumed to overestimate actual pond water concentrations because of dilution in the larger volume of the pond. In the absence of measured pond sediment concentrations, the software calculates sediment concentrations based on a conservative sediment distribution coefficient. The only available radionuclide specific concentrations for 2005 were for 129I and tritium in CFA effluents, 226Ra in INTEC effluents, 226Ra, 137Cs, and uranium isotopes in the MFC industrial waste pond ditch and sanitation lagoon, and 90Sr in TAN effluents (Table 8-7) (see Morris 2003 for a detailed description of the assessment procedure). These data were combined in a Site-wide general screening analysis. The combined sum of fractions was greater than one and failed the first screening test due to the high concentration of 226Ra.
Assuming dilution in the pond, the scenario was re-evaluated using an average concentration of radionuclides in the effluent rather than a maximum. This value (1.54) also failed the screen.
A “riparian animal” was identified as the critical organism. Although the ponds are typically lined and not attractive to riparian animals, and are surrounded by chain link fencing, it was conservatively assumed that a raccoon frequents the ponds at night approximately 50 percent of the time. The resulting estimate (0.77) passed the third screen.
For the initial terrestrial evaluation, we used maximum concentrations from the management and operating (M&O) contractor 2005 soil sampling (see Morris 2003 for a detailed description of the assessment procedure). These concentrations passed the initial screen (Table 8-8).
Based on the results of the graded approach, there is no evidence that INL-related radioactivity in soil or water is harming populations of plants or animals.
Back to top
Cahki, S. and Parks, B., 2000, CAP88-PC, Version 2.1, September.
Chew, E.W. and Mitchell, R.G., 1988, 1987 Environmental Monitoring Program Report for the Idaho National Engineering Laboratory Site, DOE/ID-12082(87), May.
Eckerman, K.F., Ryman, J.C, 1993, External Exposure to Radionuclides in Air, Water, Federal Guidance Report 12, EPA-402-R-93-081, September.
Eckerman, K.F., Wolbarst, A.B., and Richardson, A.C.B., 1988, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, Federal Guidance Report 11, EPA-520/1-88-020 September.
Environmental Protection Agency (EPA), 2002, Federal Guidance Report 13. Cancer Risk Coefficients for Environmental Exposure to Radionuclides, EPA-402-R-99-001.
Environmental Protection Agency (EPA), 2006, “National Emission Standards for Hazardous Air Pollutants,” Code of Federal Regulations, 40 CFR 61, Office of the Federal Register.
Halford, D.K., Markham, O.D., and White, G.C., 1983, “Biological Elimination of Radioisotopes by Mallards Contaminated at a Liquid Radioactive Waste Disposal Area,” Health Physics, 45: 745–756, September.
Hoff, D.L., Chew, E.W., and Rope, S.K., 1987, 1986 Environmental Monitoring Program Report for the Idaho National Engineering Laboratory Site, DOE/ID-12082(86), May.
Hoff, D.L., Mitchell, R.G., and Moore, R., 1989, 1988 Environmental Monitoring Program Report for the Idaho National Engineering Laboratory Site, DOE/ID-12082 (88), June.
ISCORS, 2004, RESRAD-BIOTA: A tool for implementing a graded approach to biota dose evaluation, ISCORS Technical Report 2004-02; DOE/EH-0676, Springfield, VA: National Technical Information Service, available from: http://homer.ornl.gov/oepa/public/bdac/.
Lewellen, W.S., Sykes, R.I., Parker, S.F., and Kornegay, F.C., 1985, Comparison of the 1981 INEL Dispersion Data with Results from a Number of Different Models, NUREG/CR-4159, U.S. Nuclear Regulatory Commission, Washington, D.C.
Markham, O.D., Halford, D.K., Autenrieth, R.E., and Dickson, R.L., 1982, “Radionuclides in Pronghorn Resulting from Nuclear Fuel Reprocessing and Worldwide Fallout,” Journal of Wildlife Management, Vol. 46, No. 1, January.
Morris, R.C., 2003, Biota Dose Assessment Guidance for the INL, NW‑ID‑2003-062, September.
Sagendorf, J.F., and Fairobent, J.E., 1986, Appraising Atmospheric Transport and Diffusion Models for Emergency Response Facilities, NUREG/CR-4603, U.S. Nuclear Regulatory Commission, Washington, D.C., May.
Sagendorf, J.F., Carter, R.G., and Clawson, K.L., 2001, MDIFFF Transport and Diffusion Model, NOAA Air Resources Laboratory, NOAA Technical Memorandum OAR ARL‑238, February.
Start, G.E., Cate, J.H., Sagendorf J.F., Ackerman, G.R., Dickson, C.R., Hukari, N.H., and Thorngren, L.G., 1985, 1981 Idaho Field Experiment, Volume 3, Comparison of Trajectories, Tracer Concentration Patterns and MESODIF Model Calculations, NUREG/CR-3488,Vol. 3, U.S. Nuclear Regulatory Commission, Washington, D.C., February.
U.S. Department of Energy Idaho Operations Office (DOE-ID), 2006, National Emissions Standards for Hazardous Air Pollutants – Calendar Year 2005 INL Report for Radionuclides, DOE/ID‑10890(06), June.
U.S. Department of Energy (DOE), 2003, “Environmental Protection Program,” DOE Order 450.1, January.
U.S. Department of Energy (DOE), 2002, A Graded Approach for Evaluating Radiation Doses to Aquatic and Terrestrial Biota, DOE-STD-1153-2002, Washington, D.C., U.S. Department of Energy, available from http://homer.ornl.gov/oepa/ public/bdac/ .
U.S. Department of Energy (DOE), 2004, “RESRAD-BIOTA: A Tool for Implementing a Graded Approach to Biota Dose Evaluation,” DOE/EH-0676, January.
U.S. Department of Energy, 1993, “Radiation Protection of the Public and the Environment,” DOE Order 5400.5, January.
U.S. Nuclear Regulatory Commission, 1977, Regulatory Guide 1.109 Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1, NRC 1.109, Revision 1, 1977.
Warren, R.W., Majors, S.J., and Morris, R.C., 2001, Waterfowl Uptake of Radionuclides from the TRA Evaporation Ponds and Potential Dose to Humans Consuming Them, Stoller-ESER‑01-40, October.
Back to top
Back to index