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Historical activities at facilities producing nuclear materials for weapons released radioactivity into the air and water. Past studies in the United States have evaluated the release, atmospheric transport and environmental accumulation of 131I from the nuclear facilities at Hanford in Washington State and the resulting dose to members of the public (Farris et al. 1994). A multi-year dose reconstruction effort (Mokrov et al. 2004) is also being conducted to produce representative dose estimates for members of the public living near Mayak, Russia, from atmospheric releases of 131I at the facilities of the Mayak Production Association. The approach to calculating individual doses to members of the public from historical releases of airborne 131I has the following general steps: • Construct estimates of releases 131I to the air from production facilities. • Model the transport of 131I in the air and subsequent deposition on the ground and vegetation. • Model the accumulation of 131I in soil, water and food products (environmental media). • Calculate the dose for an individual by matching the appropriate lifestyle and consumption data for the individual to the concentrations of 131I in environmental media at their residence location. A number of computer codes were developed to facilitate themore » study of airborne 131I emissions at Hanford. Of particular interest is DESCARTES code that modeled accumulation of 131I in environmental media (Miley et al. 1994). In addition, the CIDER computer code estimated annual doses to individuals (Eslinger et al. 1994) using the equations and parameters specific to Hanford (Snyder et al. 1994). Several of the computer codes developed to model 131I releases from Hanford are general enough to be used for other facilities. Additional codes have been developed, including the new individual dose code CiderF (Eslinger and Napier 2013), and applied to historical releases of 131I from Mayak. This document provides a data package that identifies computer code runs and associated input and output files prepared for the purpose of calculating doses to members of the public from atmospheric releases of 131I at the Mayak Production Association for the time period 1948 through 1972.« less

Historical activities at facilities producing nuclear materials for weapons released radioactivity into the air and water. Past studies in the United States have evaluated the release, atmospheric transport and environmental accumulation of 131I from the nuclear facilities at Hanford in Washington State and the resulting dose to members of the public (Farris et al. 1994). A multi-year dose reconstruction effort (Mokrov et al. 2004) is also being conducted to produce representative dose estimates for members of the public living near Mayak, Russia, from atmospheric releases of 131I at the facilities of the Mayak Production Association. The approach to calculating individual doses to members of the public from historical releases of airborne 131I has the following general steps: • Construct estimates of releases 131I to the air from production facilities. • Model the transport of 131I in the air and subsequent deposition on the ground and vegetation. • Model the accumulation of 131I in soil, water and food products (environmental media). • Calculate the dose for an individual by matching the appropriate lifestyle and consumption data for the individual to the concentrations of 131I in environmental media at their residence location. A number of computer codes were developed to facilitate themore » study of airborne 131I emissions at Hanford. The RATCHET code modeled movement of 131I in the atmosphere (Ramsdell Jr. et al. 1994). The DECARTES code modeled accumulation of 131I in environmental media (Miley et al. 1994). The CIDER computer code estimated annual doses to individuals (Eslinger et al. 1994) using the equations and parameters specific to Hanford (Snyder et al. 1994). Several of the computer codes developed to model 131I releases from Hanford are general enough to be used for other facilities. This document provides user instructions for computer codes calculating doses to members of the public from atmospheric 131I that have two major differences from the Hanford modeling sequence. First, the air transport code HYSPLIT (Draxler et al. 2012) is used instead of the RATCHET code. Second, the new individual dose code CiderF replaces the older CIDER code and five auxiliary codes.« less

This brief report documents the selection of parameters needed to support individual-dose calculations from 131I released into the environment with gaseous effluents from the Mayak Production Association.

Recent developments in evaluation of dose-response models in light of uncertain dose data (Stram and Kopecky 2003; Schafer and Gilbert 2006) have highlighted the importance of different types of uncertainties in the development of individual dose estimates. These include uncertain parameters that may be either shared or unshared within the dosimetric cohort, and also the nature of the type of uncertainty as either classical or Berkson. This report is an initial attempt to identify the nature of the various input parameters and calculational methods incorporated in the Techa River Dosimetry System (based on the TRDS-2000 implementation as a starting point, with additions for recently-developed capabilities). This report reviews the database, equations, and input parameters, and then identifies the author's interpretations of their general nature. It closes with some questions for the users of the data (epidemiologists and biostatisticians), so that the next implantation of the TRDS will provide the most useful information.

Scientists from the Environmental Science and Research Foundation, Lockheed Martin Idaho Technologies Company, the US Geological Survey, and other INEEL contractors monitored the environment on and around the INEEL to find contaminants attributable to the INEEL. During 1997, exposures from the INEEL to the public were found to be negligible. Pathways by which INEEL contaminants might reach people were monitored. These included air, precipitation, water, locally grown food (wheat, milk, potatoes, and lettuce), livestock, game animals, and direct radiation. Results from samples collected to monitor these pathways often contain radioactivity from natural sources and nuclear weapons testing carried out in the 1950s and 1960s, termed ''background radioactivity.'' According to the results obtained in 1997, radioactivity from operations at the INEEL could not be distinguished from this background radioactivity in the regions surrounding the INEEL. Because radioactivity from t! he INEEL wa s not detected by offsite environmental surveillance methods, computer models were used to estimate a radiation dose to people. The hypothetical maximum individual dose from the INEEL was calculated to be 0.03 millirem. That is 0.008 percent of an average person's annual dose from background radiation in southeast Idaho.

A preliminary radiological dose assessment of equipment decontamination, subsurface disposal, landspreading, equipment smelting, and equipment burial was conducted to address concerns regarding the presence of naturally occurring radioactive materials (NORM) in production waste streams. The assessment estimated maximum individual dose equivalents for workers and the general public. Sensitivity analyses of certain input parameters also were conducted. On the basis of this assessment, it is concluded that (1) regulations requiring workers to wear respiratory protection during equipment cleaning operations are likely to result in lower worker doses, (2) underground injection and downhole encapsulation of NORM wastes present a negligible risk to the general public, and (3) potential doses to workers and the general public related to smelting NORM-contaminated equipment can be controlled by limiting the contamination level of the initial feed. It is recommended that (1) NORM wastes be further characterized to improve studies of potential radiological doses; (2) states be encouraged to permit subsurface disposal of NORM more readily, provided further assessments support this study; results; (3) further assessment of landspreading NORM wastes be conducted; and (4) the political, economic, sociological, and nonradiological issues related to smelting NORM-contaminated equipment be studied to fully examine the feasibility of this disposal option.

Population and individual radiation dose commitments have been estimated from reported radionuclide releases from commercial power reactors operating during 1992. Fifty-year dose commitments for a 1-year exposure from both liquid and atmospheric releases were calculated for four population groups (infant, child, teenager, and adult) residing between 2 and 80 km from each of 72 reactor sites. This report tabulates the results of these calculations, showing the dose commitments for both water and airborne pathways for each age group and organ. Also included for each of the sites is an estimate of individual doses, which are compared with 10 CFR Part 50, Appendix I, design objectives. The total collective dose commitments (from both liquid and airborne pathways) for each site ranged from a high of 3.7 person-rem to a low of 0.0015 person-rem for the sites with plants in operation and producing power during the year. The arithmetic mean was 0.66 person-rem. The total population dose for all sites was estimated at 47 person-rem for the 130-million people considered at risk. The individual dose commitments estimated for all sites were below the 10 CFR 50, Appendix I, design objectives.

Population and individual radiation dose commitments have been estimated from reported radionuclide releases from commercial power reactors operating during 1991. Fifty-year dose commitments for a one-year exposure from both liquid and atmospheric releases were calculated for four population groups (infant, child, teenager and adult) residing between 2 and 80 km from each of 72 reactor sites. This report tabulates the results of these calculations, showing the dose commitments for both water and airborne pathways for each age group and organ. Also included for each of the sites is an estimate of individual doses which are compared with 10 CFR Part 50, Appendix 1 design objectives. The total collective dose commitments (from both liquid and airborne pathways) for each site ranged from a high of 22 person-rem to a low of 0.002 person-rem for the sites with plants in operation and producing power during the year. The arithmetic mean was 1.2 person-rem. The total population dose for all sites was estimated at 88 person-rem for the 130 million people considered at risk. The individual dose commitments estimated for all sites were below the Appendix 1 design objectives.

This report analyzes the radionuclide dose impact of the PUREX process waste stream. The radionuclide ingestion and inhalation pathways are analyzed. Two spent fuel assemblies processed in the Separation facilities are analyzed, the Mark 31A and Mark 31B. The individual radionuclide significance to dose is evaluated in terms of dose percentage. Comparing the radionuclide individual dose value allows the determination of those radionuclides whose dose impact is significant. The results of this analysis demonstrate that a limited number of radionuclides contribute 1% or more to the total dose and that the major contributor to the sludge source dose is strontium. The results obtained permit reducing the list of radionuclides to be considered in the development of source terms to support the High Level Waste Safety Analysis Report.

An EXCEL{reg_sign} spreadsheet has been developed that, when combined with the PC version of XOQDOQ, will generate estimates of maximum individual dose from routine atmospheric releases of radionuclides. The spreadsheet, MAXINE, utilizes a variety of atmospheric dispersion factors to calculate radiation dose as recommended by the US Nuclear Regulatory Commission in Regulatory Guide 1.109 [USNRC 1977a]. The methodology suggested herein includes use of both the MAXINE spreadsheet and the PC version of XOQDOQ.

The offsite individual residing at the SRS boundary location of maximum exposure (maximum individual) received a dose from SRS atmospheric releases in 1992 of 0.10 mrem. Tritium oxide releases were responsible for more than 90% of the atmospheric dose. Ninety-nine percent of the atmospheric dose is accounted for with the inclusion of I-129, U-235, Pu-238 and Pu-239. The 50-mile population received a collective dose of 6.4 person-rem from 1992 releases. Atmospheric radiation dose is dominated by the inhalation and vegetation consumption pathways. The maximum dose received by an offsite individual as a result of SRS liquid releases in 1992 was 0.13 mrem. This dose is dominated by Cs-137 that is accumulated in Savannah River fish and tritium in drinking water. More than 99% of the maximum individual dose from liquid releases results from Cs-137, tritium, Sr-90, and Pu-239. The population dose from liquid releases in 1992 was 2.5 person-rem. The major exposure pathway to the population is drinking water. Radiation dose to the general public from operations at the Savannah River Site continues to be a very small fraction of the natural background dose. A resident of the CSRA receives about 300 mrem per year from background radiation. The populationmore » within 50-miles of the SRS (620,000) and at the downstream water treatment facilities (65,000), therefore, receives a natural background population dose of approximately 200,000 person-rem.« less

This report summarizes the occupational radiation exposure information that has been reported to the NRC`s Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1992. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10CFR20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10CFR20.408. The 1992 annual reports submitted by about 364 licensees indicated that approximately 204,365 individuals were monitored, 183,927 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.16 rem (cSv) and an average measurable dose of about 0.30 (cSv). Termination radiation exposure reports were analyzed to reveal that about 74,566 individuals completed their employment with one or more of the 364 covered licensees during 1992. Some 71,846 of these individuals terminated from power reactor facilities, and about 9,724 of them were considered to be transient workers who received an average dose of 0.50 rem (cSv).

This Software Development Plan (SDP) outlines all software activities required to obtain functional environmental accumulation and individual dose codes for the Hanford Environmental Dose Reconstruction (HEDR) project. The modeling activities addressed use the output of the air transport-code HATCHET to compute radionuclide concentrations in environmental pathways, and continue on through calculations of dose for individuals. The Hanford Environmental Dose Reconstruction (HEDR) Project has a deliverable in the June 1993 time frame to be able to start computing doses to individuals from nuclear-related activities on the Hanford Site during and following World War II. The CIDER code will compute doses and their uncertainties for individuals living in the contaminated environment computed by DESCARTES. The projected size of the code is 3000 lines.

As part of the Hanford Environmental Dose Reconstruction (HEDR) Project, Battelle, Pacific Northwest Laboratories reviewed literature and data on radionuclide concentrations and distribution in the water, sediment, and biota of the Columbia River and adjacent coastal areas. Over 600 documents were reviewed including Hanford reports, reports by offsite agencies, journal articles, and graduate theses. Radionuclide concentration data were used in preliminary estimates of individual dose for the period 1964 through 1966. This report summarizes the literature and database reviews and the results of the preliminary dose estimates.

As part of the Hanford Environmental Dose Reconstruction (HEDR) Project, Battelle, Pacific Northwest Laboratories reviewed literature and data on radionuclide concentrations and distribution in the water, sediment, and biota of the Columbia River and adjacent coastal areas. Over 600 documents were reviewed including Hanford reports, reports by offsite agencies, journal articles, and graduate theses. Radionuclide concentration data were used in preliminary estimates of individual dose for the period 1964 through 1966. This report summarizes the literature and database reviews and the results of the preliminary dose estimates.

Preliminary radiation doses were estimated and reported during Phase I of the Hanford Environmental Dose Reconstruction (HEDR) Project. As the project has progressed, additional information regarding the magnitude and timing of past radioactive releases has been developed, and the general scope of the required calculations has been enhanced. The overall HEDR computational model for computing doses attributable to atmospheric releases from Hanford Site operations is called HEDRIC (Hanford Environmental Dose Reconstruction Integrated Codes). It consists of four interrelated models: source term, atmospheric transport, environmental accumulation, and individual dose. The source term and atmospheric transport models are documented elsewhere. This report describes the initial implementation of the design specifications for the environmental accumulation model and computer code, called DESCARTES (Dynamic EStimates of Concentrations and Accumulated Radionuclides in Terrestrial Environments), and the individual dose model and computer code, called CIDER (Calculation of Individual Doses from Environmental Radionuclides). The computations required of these models and the design specifications for their codes were documented in Napier et al. (1992). Revisions to the original specifications and the basis for modeling decisions are explained. This report is not the final code documentation but gives the status of the model and code development to date. Final code documentation is scheduled to be completed in FY 1994 following additional code upgrades and refinements. The user's guide included in this report describes the operation of the environmental accumulation and individual dose codes and associated pre- and post-processor programs. A programmer's guide describes the logical structure of the programs and their input and output files.

The water level of Par Pond has been lowered over the past several months to reduce the effects in the event of catastrophic dam failure while assessing the condition of the dam and determining if repairs are necessary. In lowering the level of Par Pond, 60 billion liters of water containing low levels of tritium and cesium-137 were discharged to several onsite streams. SRS surface streams flow to the Savannah River. An assessment made to determine the total amount of tritium and Cs-137 discharged and to estimate the consequences to downstream Savannah River users. It is estimated that a total of 160 curies of tritium were displaced from Par Pond to the Savannah River between June 28, 1991 and September 19, 1991. This release could hypothetically result in a maximum individual dose of 3. 2{times}10{sup {minus}4} mrem and a total (80-km and drinking water populations) population dose of 1.4{times}10{sup {minus}2} person-rem. Likewise, a maximum individual dose of 5.0{times}10{sup {minus}2} mrem and a total population dose of 1.7{times}10{sup {minus}1} person- rem are predicted as a result of an estimated 0.21 curies of Cs-137 being discharged from Par Pond to the Savannah River.

Radiological dose to the offsite maximum individual or the 50-mile population is often estimated assuming that operational atmospheric releases originate from the geographical center of the Savannah River Site. Historically, meteorological data collected from instrumentation on the H-Area tower have been utilized to estimate atmospheric dispersion from centrally located ``releases.`` This paper examines the effect on dose predictions using meteorological data from seven onsite towers located at A, C, D, F, H, K, and P areas to describe meteorological conditions at the central release location. Maximum individual dose estimates using both annual and short-term average air concentrations are well within 20% of the mean prediction. Population dose estimates are also within 20% of the mean prediction. Population dose estimates are also within 20% of the mean except for two-hour doses using H-Area meteorological data, which is less than the average estimate by about 30%.

Population radiation dose commitments have been estimated from reported radionuclide releases from commercial power reactors operating during 1977. Fifty-year dose commitments from a one-year exposure were calculated from both liquid and atmospheric releases for four population groups (infant, child, teen-ager and adult) residing between 2 and 80 km from each site. This report tabulates the results of these calculations, showing the dose commitments for both liquid and airborne pathways for each age group and organ, Also included for each site is a histogram showing the fraction of the total population within 2 to 80 km around each site receiving various average dose commitments from the airborne pathways. The total dose commitment from both liquid and airborne pathways ranged from a high of 220 person-rem to a low of 0.003 person-rem with an arithmetic mean of 16 person-rem. The total population dose for all sites was estimated at 700 person-rem for the 92 million people considered at risk. The average individual dose commitment from all pathways on a site basis ranged from a low of 2 x 10{sup -5} mrem to a high of 0.1 mrem. No attempt was made in this study to determine the maximum dose commitment received bymore » any one individual from the radionuclides released at any of the sites.« less

Population radiation dose commitments have been estimated from reported radionuclide releases from commercial power reactors operating during 1979. Fifty-year dose commitments from a one-year exposure were calculated from both liquid and atmospheric releases for four population groups (infant, child, teen-ager and adult) residing between 2 and 80 km from each site. This report tabulates the results of these calculations, showing the dose commitments for both liquid and airborne pathways for each age group and organ. Also included for each site is a histogram showing the fraction of the total population within 2 to 80 km around each site receiving various average dose commitments from the airborne pathways. The total dose commitment from both liquid and airborne pathways ranged from a high of 1300 person-rem to a low of 0.0002 person-rem with an arithmetic mean of 38 person-rem. The total population dose for all sites was estimated at 1800 person-rem for the 94 million people considered at risk. The average individual dose commitment from all pathways on a site basis ranged from a low of 2 x 10/sup -6/ mrem to a high of 0.7 mrem. No attempt was made in this study to determine the maximum dose commitment received by any one individual from the radionuclides released at any of the sites.

A discussion of the offsite effective dose equivalents reported for 1988 SRS releases serves as the basis for this report. Detailed analyses of atmospheric and liquid release trends and their consequences in terms of relative importance among facilities, radionuclides, and exposure pathways have also been included. Releases of radioactivity to the atmosphere were generally lower in 1988 than in 1987. No major unplanned tritium releases were recorded during the year. However, there were three inadvertent releases of Pu-238 from F Area in January, March and October of 160, 32 and 83 uCi, respectively. Radioactive releases to onsite streams from direct discharges and seepage basin migration decreased in 1988. However, as a result of a decrease in the flow rate of the Savannah River in 1988, higher offsite doses were reported. The maximum individual dose, conversely, was down from 1987. This decrease reflected the fact that the maximum individual dose is most significantly affected by the cesium concentration in fish. In terms of largest contributors to dose, the releases were dominated by tritium, Cs-137 and to a much lesser extent Sr-90. With respect to the offsite population, doses from atmospheric releases are generally higher than those from liquid releases, and thismore » trend continued in 1988. Analyses of 1988 data and the data for the preceding decade suggest that radioactive releases from the SRS during this period have not significantly impacted the offsite population.« less

The Radiation Exposure Information Reporting System (REIRS) was initiated by the Atomic Energy Commission (AEC) in 1968. While the system has provided a general overview of radiation exposures associated with AEC/ERDA/DOE operations and has satisfied the original intent for a central information system, the need for more detailed information has become evident. The alternatives addressed for a radiation exposure information system were no change in current system, clarification of DOE Order for current system, increased summary information from sites, centralized annual individual dose (exposure) system, and annual dose summary and locator files. A majority of the DOE Ad Hoc Committee has concurred to recommend the annual dose summary and locator files (ADSLF). The acceptance of the ADSLF alternative as the DOE-wide radiation exposure system would give DOE added capability and flexibility in responding to requests for information and would reduce the impact on the sites of special survey requests.

Three different dosimeters were used to calibrate Savannah River Plant /sup 60/Co sources having intensities greater than 10/sup 7/ rads/hr. These dosimeters are (a) ceric sulfate dissolved in 0.4M H/sub 2/SO/sub 4/, (b) oxalic acid dissolved in water, and (c) a commercially available nylon film containing a radiochromic dye. Response per unit dose to these dosimeters is independent of radiation intensity at 10/sup 4/ to 10/sup 11/ rads/hr. The dosimeters were calibrated at 6.0 x 10/sup 5/ rads/hr with a /sup 60/Co source whose intensity was determined with the standard Fricke dosimeter. For the sources at 10/sup 7/ rads/hr or greater, intensities were calculated from slopes of linear plots of dosimeter response versus irradiation time. Individual dose rates varied from 1.0 x 10/sup 7/ to 4.6 x 10/sup 7/ rads/hr. Each source was calibrated with at least two different dosimeters. Relative standard deviations varied from 2 to 9 percent. A conservative estimate of the uncertainty in the accuracy of these dosimeters is 10 percent. Of the three dosimeters, the nylon film is easiest to use and is therefore recommended for future calibrations.

This report provides criteria for evaluating the response of personnel dosimeters to radiation at nonperpendicular incidence. The US Department of Energy Laboratory Accreditation Program (DOELAP) ensures that dosimetry systems at DOE facilities meet acceptable standards for precision and accuracy. In the past, these standards were limited to tests for system variability, energy dependence, and level of detection. The proposed criteria will broaden the scope of DOELAP to include the angular response of personnel dosimeters. Because occupational exposures in the workplace are rarely due to radiation from only one direction, dosimeters must accurately assign individual dose equivalent from irradiation at any forward angle of incidence. Including an angular response criterion in DOELAP would improve the quality of personnel monitoring provided that the criterion is developed from appropriate dose quantities. This report provides guidance for assigning individual dose equivalents for radiation fields at nonperpendicular incidence to the dosimeter. 21 refs., 10 figs., 10 tabs.