APPENDIX E: HUMAN HEALTH E.1 Introduction Detailed information on the potential impacts to humans from the normal operational releases of radioactivity and hazardous chemicals from the tritium supply technologies and recycling facilities is presented in this appendix. This information is intended to support assessments of normal operation for the tritium supply and recycling facilities described in sections 4.2.3.9, 4.3.3.9, 4.4.3.9, 4.5.3.9, and 4.6.3.9 of this Programmatic Environmental Impact Statement (PEIS). Section E.2 provides information on radiological impacts while section E.3 provides information on hazardous chemical impacts. E.2 Radiological Impacts To Human Health Section E.2 presents supporting information on the potential radiological impacts of normal operation to humans. This section provides the reader with background information on the nature of radiation (section E.2.1), the methodology used to calculate radiological impacts (section E.2.2), radiological releases from tritium supply and recycling facilities (section E.2.3), and radiological impacts from various tritium supply technologies and recycling facilities at each site (sections E.2.4 through E.2.8). E.2.1 Background E.2.1.1 Nature of Radiation and Its Effects on Humans What is Radiation? Humans are constantly exposed to radiation from the solar system and from the earth's rocks and soil. This radiation contributes to the natural background radiation that has always surrounded us. But there are also manmade sources of radiation, such as medical and dental x-rays, household smoke detectors, and materials released from nuclear and coal-fired powerplants. All matter in the universe is composed of atoms, and radiation comes from the activity of these tiny particles. Atoms are made up of even smaller particles (protons, neutrons, and electrons). The number and arrangement of these particles distinguishes one atom from another. Atoms of different types are known as elements. There are over 100 natural and manmade elements. Some of these elements, such as uranium, radium, plutonium, and thorium, share a very important quality: they are unstable. As they change into more stable forms, invisible waves of energy or particles, known as ionizing radiation, are released. Radioactivity is the emitting of this radiation. Ionizing radiation refers to the fact that this energy force can ionize, or electrically charge atoms by stripping off electrons. Ionizing radiation can cause a change in the chemical composition of many things, including living tissue (organs), which can affect the way they function. The effects on people of radiation that is emitted during disintegration (decay) of a radioactive substance depends on the kind of radiation (alpha and beta particles, and gamma and x-rays) and the total amount of radiation energy absorbed by the body. Alpha particles are the heaviest of these direct types of ionizing radiation, and despite a speed of about 10,000 miles per second, they can travel only a few inches in the air. Alpha particles lose their energy almost as soon as they collide with anything. They can easily be stopped by a sheet of paper or the skin's surface. Beta particles are much lighter than alpha particles. They can travel as much as 100,000 miles per second and can travel in the air for a distance of about 10feet. Beta particles can pass through a sheet of paper but may be stopped by a thin sheet of aluminum foil or glass. Gamma and x-rays, unlike alpha or beta particles, are waves of pure energy. Gamma rays travel at the speed of light (186,000 miles per second). Gamma radiation is very penetrating and requires a thick wall of concrete, lead, or steel to stop it. The neutron is another particle that contributes to radiation exposure, both directly and indirectly. Indirect exposure is associated with the gamma rays and alpha particles that are emitted following neutron capture in matter. A neutron has about one quarter the weight of an alpha particle and can travel at speeds of up to 24,000 miles per second. Neutrons are more penetrating than beta particles, but less than gamma rays. The radioactivity of a material decreases with time. The time it takes a material to lose half of its original radioactivity is its half-life. For example, a quantity of iodine-131, a material that has a half-life of 8 days, will lose half of its radioactivity in that amount of time. In 8 more days, one-half of the remaining radioactivity will be lost, and so on. Eventually, the radioactivity will essentially disappear. Each radio- active element has a characteristic half-life. The half-lives of various radioactive elements may vary from millionths of a second to millions of years. As a radioactive element gives up its radioactivity, it often changes to an entirely different element, one that may or may not be radioactive. Eventually, a stable element is formed. This transformation may take place in several steps and is known as a decay chain. Radium, for example, is a naturally occurring radioactive element with a half-life of 1,622 years. It emits an alpha particle and becomes radon, a radioactive gas with a half-life of only 3.8 days. Radon decays to polonium and, through a series of steps, to bismuth and ultimately to lead. Units of Radiation Measure. Scientists and engineers use a variety of units to measure radiation. These different units can be used to determine the amount, type and intensity of radiation. Just as heat can be measured in terms of its intensity or its effects using units of calories or degrees, amounts of radiation can be measured in curies, rads, or rems. The curie, named after the French scientists Marie and Pierre Curie, describes the "intensity" of a sample of radioactive material. The rate of decay of 1 gram of radium is the basis of this unit of measure. It is equal to 3.7x1010 disintegrations (decays) per second. The total energy absorbed per unit quantity of tissue is referred to as absorbed dose. The rad is the unit of measurement for the physical absorption of radiation. Much like sunlight heats the pavement by giving up an amount of energy to it, radiation gives up rads of energy to objects in its path. One rad is equal to the amount of radiation that leads to the deposition of 0.01 joule of energy per kilogram of absorbing material. A rem is a measurement of the dose from radiation based on its biological effects. The rem is used to measure the effects of radiation on the body, much like degrees Celsius can be used to measure the effects of sunlight heating pavement. Thus, 1 rem of one type of radiation is presumed to have the same biological effects as 1 rem of any other type of radiation. This standard allows comparison of the biological effects of radionuclides that emit different types of radiation. An individual may be exposed to ionizing radiation externally from a radioactive source outside the body and/or internally from ingesting radioactive material. The external dose is different from the internal dose. An external dose is delivered only during the actual time of exposure to the external radiation source. An internal dose, however, continues to be delivered as long as the radioactive source is in the body, although both radioactive decay and elimination of the radionuclide by ordinary metabolic processes decrease the dose rate with the passage of time. The dose from internal exposure is calculated over 50 years following the initial exposure. The three types of doses calculated in this PEIS include an external dose, an internal dose, and a combined external and internal dose. Each type of dose is discussed below. External Dose. The external dose can arise from several different pathways. All these pathways have in common the fact that the radiation causing the exposure is external to the body. In this PEIS, these pathways include exposure to a cloud of radiation passing over the receptor, standing on ground that is contaminated with radioactivity, swimming in contaminated water, and boating in contaminated water. The appropriate measure of dose is called the effective dose equivalent. It should be noted that if the receptor departs from the source of radiation exposure, his dose rate will be reduced. It is assumed that external exposure occurs uniformly during the year. Internal Dose. The internal dose arises from a radiation source entering the human body through either ingestion of contaminated food and water or inhalation of contaminated air. In this PEIS, pathways for internal exposure include ingestion of crops contaminated either by airborne radiation depositing on the crops or by irrigation of crops using contaminated water sources, ingestion of animal products from animals that ingested contaminated food, ingestion of contaminated water, inhalation of contaminated air, and absorption of contaminated water through the skin during swimming. Unlike external exposures, once the radiation enters the body, it remains there for various periods of time depending on decay and biological elimination rates. The unit of measure for internal doses is the committed dose equivalent. It is the internal dose that each body organ receives from 1 "year intake" (ingestion plus inhalation). Normally, a 50- or 70-year dose-commitment period is used (i.e., the 1-year intake period plus 49 or 69 years). The dose rate increases during the 1 year of intake. The dose rate, after the one year of intake, slowly declines as the radioactivity in the body continues to produce a dose. The integral of the dose rate over the 50 or 70years gives the committed dose equivalent. In this PEIS, a 50-year dose-commitment period was used. The various organs of the body have different susceptibilities to harm from radiation. The quantity that takes these different susceptibilities into account to provide a broad indicator of the risk to the health of an individual from radiation is called the committed effective dose equivalent. It is obtained by multiplying the committed dose equivalent in each major organ or tissue by a weighting factor associated with the risk susceptibility of the tissue or organ, then summing the totals. It is possible that the committed dose equivalent to an organ is larger than the committed effective dose equivalent if that organ has a small weighting factor. The concept of committed effective dose equivalent applies only to internal pathways. Differences in radionuclide characteristics lead to different internal doses. For example, for the same amount of radioactivity, in curies, taken into the body, the dose from tritium is much less than from uranium or plutonium. Tritium emits a weak beta particle and is biologically eliminated from the body over several weeks. Uranium and plutonium emit relatively high-energy alpha particles and are retained in the body for periods of several months to many years. Combined External and Internal Dose. For convenience, the sum of the committed effective dose equivalent from internal pathways and the effective dose equivalent from external pathways is also called the committed effective dose equivalent in this PEIS (note that in DOE Order 5400.5 this quantity is called the effective dose equivalent). The units used in this PEIS for committed dose equivalent, effective dose equivalent, and committed effective dose equivalent to an individual are the rem and mrem (1/1000 of 1 rem). The corresponding unit for the collective dose to a population (the sum of the doses to members of the population, or the product of the number of exposed individuals and their average dose) is the person-rem. Sources of Radiation. The average American receives a total of about 350 mrem per year from all sources of radiation, both natural and manmade. The sources of radiation can be divided into six different categories: cosmic radiation, terrestrial radiation, internal radiation, consumer products, medical diagnosis and therapy, and other sources. Each category is discussed below. Cosmic radiation is ionizing radiation resulting from energetic charged particles from space continuously hitting the earth's atmosphere. These particles and the secondary particles and photons they create are cosmic radiation. Because the atmosphere provides some shielding against cosmic radiation, the intensity of this radiation increases with altitude above sea level. For the sites considered in this PEIS, the cosmic radiation ranged from about 30 to 50 mrem per year. The average annual dose to the people in the United States is about 27 mrem. External terrestrial radiation is the radiation emitted from the radioactive materials in the earth's rocks and soils. The average annual dose from external terrestrial radiation is about 28 mrem. The external terrestrial radiation for the sites in this PEIS ranged from about 30 to 75 mrem per year. Internal radiation arises from the human body metabolizing natural radioactive material that has entered the body by inhalation or ingestion. Natural radionuclides in the body include isotopes of uranium, thorium, radium, radon, polonium, bismuth, potassium, rubidium, and carbon. The major contributor to the annual dose equivalent for internal radioactivity are the short-lived decay products of radon which contribute about 200 mrem per year. The average dose from other internal radionuclides is about 39mrem per year. Consumer products also contain sources of ionizing radiation. In some products, like smoke detectors and airport x-ray machines, the radiation source is essential to the products' operation. In other products, such as television and tobacco, the radiation occurs incidentally to the product function. The average annual dose is about 10 mrem. Radiation is an important diagnostic medical tool and cancer treatment. Diagnostic x-rays result in an average annual exposure of 39 mrem. Nuclear medical procedures result in an average annual exposure of 14 mrem. There are a few additional sources of radiation that contribute minor doses to individuals in the United States. The dose from nuclear fuel cycle facilities, such as uranium mines, mills and fuel processing plants, nuclear power plants, and transportation routes has been estimated to be less than 1 mrem per year. Radioactive fallout from atmospheric atomic bomb tests, emissions of radioactive material from Department of Energy (DOE) facilities, emissions from certain mineral extraction facilities, and transportation of radioactive materials contributes less than 1 mrem per year to the average dose to an individual. Air travel contributes approximately 1 mrem per year to the average dose. The collective (or population) dose to an exposed population is calculated by summing the estimated doses received by each member of the exposed population. This total dose received by the exposed population is measured in person-rem. For example, if 1,000 people each received a dose of 1 millirem (0.001 rem), the collective dose is (1,000 persons x 0.001 rem = 1.0 person-rem). Alternatively, the same collective dose (1.0 person-rem) results from 500people each of whom received a dose of 2millirem (500 persons x 2 millirem = 1 person-rem). Limits of Radiation Exposure. The amount of manmade radiation that the public may be exposed to is limited by Federal regulations. Although most scientists believe that radiation absorbed in small doses over several years is not harmful, U.S. Government regulations assume that the effects of all radiation exposures are cumulative. The exposure to a member of the general public from DOE facility releases into the atmosphere is limited by the Environmental Protection Agency (EPA) to an annual dose of 10 mrem, in addition to the natural background and medical radiation normally received (40 CFR 61, Subpart H). DOE also limits to 10mrem, the dose annually received from material released into the atmosphere (DOE Order 5400.5). EPA and DOE also limit the annual dose to the general public from radioactive releases to drinking water to 4 mrem (40 CFR 141; DOE Order 5400.5). The annual dose from all radiation sources from a site is limited by the EPA to 25 mrem (40 CFR 190). The DOE annual limit of radiation dose to a member of the general public from all DOE facilities is 100mrem total from all pathways (DOE Order 5400.5). For people working in an occupation that involves radiation, DOE and the Nuclear Regulatory Commission (NRC) limit doses to 5 rem (5,000mrem) in any one year (10 CFR 20; 10 CFR 835). E.2.1.2 Health Effects Radiation exposure and its consequences are topics of interest to the general public. For this reason, this PEIS places much emphasis on the consequences of exposure to radiation, even though the effects of radiation exposure under most circumstances evaluated in this PEIS are small. This section explains the basic concepts used in the evaluation of radiation effects in order to provide the background for later discussion of impacts. Radiation can cause a variety of ill-health effects in people. The most significant ill-health effect to depict the consequences of environmental and occupational radiation exposure is induction of cancer fatalities. This effect is referred to as "latent" cancer fatalities because the cancer may take many years to develop and for death to occur and may not actually be the cause of death. In the discussions that follow, it should be noted that all fatal cancers are latent and the term "latent" is not used. Health impacts from radiation exposure, whether from sources external or internal to the body, generally are identified as "somatic" (affecting the individual exposed) or "genetic" (affecting descendants of the exposed individual). Radiation is more likely to produce somatic effects rather than genetic effects. Therefore, for this PEIS, only the somatic risks are presented. The somatic risks of most importance are the induction of cancers. Except for leukemia, which can have an induction period (time between exposure to carcinogen and cancer diagnosis) of as little as 2 to 7 years, most cancers have an induction period of more than 20 years. For a uniform irradiation of the body, the incidence of cancer varies among organs and tissues; the thyroid and skin demonstrate a greater sensitivity than other organs. However, such cancers also produce relatively low mortality rates because they are relatively amenable to medical treatment. Because of the readily available data for cancer mortality rates and the relative scarcity of prospective epidemiologic studies, somatic effects leading to cancer fatalities rather than cancer incidence are presented in this PEIS. The numbers of cancer fatalities can be used to compare the risks among the various alternatives. The fatal cancer risk estimators presented in this appendix for radiation technically apply only to low-Linear Energy Transfer radiation (gamma rays and beta particles). However, on a per rem rather than a per rad basis, the fatal risk estimators are higher for this type of radiation than for high-Linear Energy Transfer radiation (alpha particles). In this PEIS, the low-Linear Energy Transfer risk estimators are con- servatively assumed to apply to all radiation exposures. The National Research Council's Committee on the Biological Effects of Ionizing Radiations has prepared a series of reports to advise the U.S. Government on the health consequences of radiation exposures. The latest of these reports, Health Effects of Exposure to Low Levels of Ionizing Radiation BEIR V, published in 1990, provides the most current estimates for excess mortality from leukemia and cancers other than leukemia expected to result from exposure to ionizing radiation. The BEIR V report updates the models and risk estimates provided in the earlier report of the BEIR III Committee, The Effects of Populations of Exposure to Low-Levels of Ionizing Radiation, published in 1980. BEIR V models were developed for application to the U.S. population. BEIR V provides estimates that are consistently higher than those in BEIR III. This is attributed to several factors, including the use of a linear dose response model for cancers other than leukemia, revised dosimetry for the Japanese atomic bomb survivors, and additional follow up studies of the atomic bomb survivors and other cohorts. BEIR III employs constant relative and absolute risk models, with separate coefficients for each sex and several age-at-exposure groups, while BEIR V develops models in which the excess relative risk is expressed as a function of age at exposure, time after exposure, and sex for each of several cancer categories. BEIR III models were based on the assumption that absolute risks are comparable between the atomic bomb survivors and the U.S. population, while BEIR V models were based on the assumption that the relative risks are comparable. For a disease such as lung cancer, where baseline risks in the United States are much larger than those in Japan, the BEIR V approach leads to larger risk estimates than the BEIRIII approach. The models and risk coefficients in BEIR V were derived through analyses of relevant epidemiologic data including the Japanese atomic bomb survivors, ankylosis spondylitis patients, Canadian and Massachusetts fluoroscopy patients (breast cancer), New York postpartum mastitis patients (breast cancer), Israel tinea capitis patients (thyroid cancer), and Rochester thymus patients (thyroid cancer). Models for leukemia, respiratory cancer, digestive cancer, and other cancers used only the atomic bomb survivor data, although results of analyses of the ankylosis spondylitis patients were considered. Atomic bomb survivor analyses were based on revised dosimetry with an assumed Relative Biologi- cal Effectiveness of 20 for neutrons and were restricted to doses of less than 400 rads. Estimates of risks of fatal cancers other than leukemia were obtained by totaling the estimates for breast cancer, respiratory cancer, digestive cancer, and other cancers. Risk Estimates for Doses Received During an Accident. BEIR V includes risk estimates for a single exposure of 10 rem to a population of 100,000people (106 person-rem). In this case, fatality estimates for leukemia, breast cancer, respiratory cancer, digestive cancer, and other cancers are given for both sexes and nine age-at-exposure groups. These estimates, based on the linear model, are summarized in table E.2.1.2-1. The average risk estimate from all ages and both sexes is 885 excess cancer fatalities per million person-rem. This value has been conservatively rounded up to 1,000 excess cancer fatalities per million person-rem. Although values for other health effects are not presented in this PEIS, the risk estimators for nonfatal cancers and for genetic disorders to future generations are estimated to be approximately 200 and 260 per million person-rem, respectively. These values are based on information presented in ICRP Publication 60 and are seen to be 20 percent and 26percent, respectively, of the fatal cancer estimator. Thus, for example, if the number of excess fatal cancers is projected to be "x," the number of excess genetic disorders would be 0.26 times "x." Risk Estimates for Doses Received During Normal Operation. For low doses and dose rates, a linear-quadratic model was found to provide a significantly better fit to the data for leukemia than a linear one, and leukemia risks were based on a linear-quadratic function. This reduces the effects by a factor of two over estimates that are obtained from the linear model. For other cancers, linear models were found to provide an adequate fit to the data, and were used for extrapolation to low doses. However, the BEIR V Committee recommended reducing these linear estimates by a factor between 2 and 10 for doses received at low dose rates. For this PEIS, a risk reduction factor of two was adopted for conservatism. Table E.2.1.2-1.-Lifetime Risks per 100,000 Persons Exposed to a Single Exposure of 10 Rem Gender Type of Fatal Cancer - Leukemia Cancers Total Other Than Cancers Leukemia Male 220 660 880 Female 160 730 890 Average 190 695 885 Based on the above discussion, the resulting risk estimator would be equal to half the value observed for accident situations or approximately 500 excess fatal cancer per million person-rem (0.0005 excess fatal cancer per person-rem). This is the risk value used in this PEIS to calculate fatal cancers to the general public during normal operations. For workers, a value of 400 excess fatal cancers per million person-rem (0.0004 excess fatal cancer per person-rem) is used in this PEIS. This lower value reflects the absence of children in the workforce. Again, based on information provided in ICRP Publication 60, the health risk estimators for nonfatal cancers and genetic disorders among the public are 20 percent and 26 percent, respectively, of the fatal cancer risk estimator. For workers they are both 20percent of the fatal cancer risk estimator. For this PEIS, only fatal cancers are presented. The risk estimates may be applied to calculate the effects of exposing a population to radiation. For example, in a population of 100,000 people exposed only to natural background radiation (0.3 rem per year), 15 latent cancer fatalities per year would be inferred to be caused by the radiation (100,000persons x 0.3 rem per year x 0.0005 latent cancer fatalities per person-rem = 15 latent cancer fatalities per year). Sometimes, calculations of the number of excess cancer fatalities associated with radiation exposure do not yield whole numbers and, especially in environmental applications, may yield numbers less than 1.0. For example, if a population of 100,000 were exposed as above, but to a total dose of only 0.001rem, the collective dose would be 100 person-rem, and the corresponding estimated number of latent cancer fatalities would be 0.05 (100,000persons x 0.001 rem x 0.0005 latent cancer fatalities/person-rem = 0.05 latent fatal cancers). How should one interpret a nonintegral number of latent cancer fatalities such as 0.05? The answer is to interpret the result as a statistical estimate. That is, 0.05 is the average number of deaths that would result if the same exposure situation were applied to many different groups of 100,000 people. In most groups, no person (0 people) would incur a latent cancer fatality from the 0.001 rem dose each member would have received. In a small fraction of the groups, 1 latent fatal cancer would result; in exceptionally few groups, 2 or more latent fatal cancers would occur. The average number of deaths over all the groups would be 0.05 latent fatal cancers (just as the average of 0, 0, 0, and 1 is 1/4, or 0.25). The most likely outcome is 0 latent cancer fatalities. These same concepts apply to estimating the effects of radiation exposure on a single individual. Consider the effects, for example, of exposure to background radiation over a lifetime. The "number of latent cancer fatalities" corresponding to a single individual's exposure over a (presumed) 72-year lifetime to 0.3 rem per year is the following: 1 person x 0.3 rem per year x 72 years x 0.0005 latent cancer fatalities per person-rem = 0.011 latent cancer fatalities. Again, this should be interpreted in a statistical sense; that is, the estimated effect of background radiation exposure on the exposed individual would produce a 1.1-percent chance that the individual might incur a latent fatal cancer caused by the exposure. Presented another way, this method estimates that approximately 1.1 percent of the population might die of cancers induced by the radiation background. E.2.2 Methodology for Estimating Radiological Impacts of Normal Operation The radiological impacts of normal operation of reactors and support facilities were calculated using Version 1.485 of the GENII computer code. Site-specific and technology-specific input data were used, including location, meteorology, population, food production and consumption, and source terms. The GENII code was used for analysis of normal operations and design basis accidents. Section E.2.2.1 briefly describes GENII and outlines the approach used for normal operations. The approach used for design basis accidents is discussed in appendix F. E.2.2.1 GENII Computer Code The GENII computer model, developed by Pacific Northwest Laboratory for the U.S. Department of Energy, is an integrated system of various computer modules that analyze environmental contamination resulting from acute or chronic releases to, or initial contamination in, air, water, or soil. The model calculates radiation doses to individuals and populations. The GENII computer model is well documented for assumptions, technical approach, methodology, and quality assurance issues (GENII - The Hanford Environmental Radiation Dosimetry Software System (December 1988)). The GENII computer model has gone through extensive quality assurance and quality control steps. These include the comparison of results from model computations against those from hand calculations, and the performance of internal and external peer reviews. Recom- mendations given in these reports were incorporated into the final GENII computer model, as deemed appropriate. For this PEIS only the ENVIN, ENV, and DOSE computer modules were used. The codes are connected through data transfer files. The output of one code is stored in a file that can be used by the next code in the system. In addition, a computer code called CREGENII was prepared to aid the user with the preparation of input files into GENII. CREGENII. The CREGENII code helps the user, through a series of interactive menus and questions, prepare a text input file for the environmental dosimetry programs. In addition, CREGENII prepares a batch processing file to manage the file handling needed to control the operations of subsequent codes and to prepare an output report. ENVIN. The ENVIN module of the GENII code controls the reading of the input files prepared by CREGENII and organizes the input for optimal use in the environmental transport and exposure module, ENV. The ENVIN code interprets the basic input, reads the basic GENII data libraries and other optional input files, and organizes the input into sequential segments on the basis of radionuclide decay chains. A standardized file that contains scenario, control, and inventory parameters is used as input to ENVIN. Radionuclide inventories can be entered as functions of releases to air or water, concentrations in basic environmental media (air, soil, or water), or concen- trations in foods. If certain atmospheric dispersion options have been selected, this module can generate tables of atmospheric dispersion parameters that will be used in later calculations. If the finite plume air submersion option is requested in addition to the atmospheric dispersion calculations, preliminary energy-dependent finite plume dose factors also are prepared. The ENVIN module prepares the data transfer files that are used as input by the ENV module; ENVIN generates the first portion of the calculation documentation-the run input parameters report. ENV. The ENV module calculates the environmental transfer, uptake, and human exposure to radionuclides that result from the chosen scenario for the user-specified source term. The code reads the input files from ENVIN and then, for each radionuclide chain, sequentially performs the precalculations to establish the conditions at the start of the exposure scenario. Environmental concentrations of radionuclides are established at the beginning of the scenario by assuming decay of preexisting sources, considering biotic transport of existing subsurface contamination, and defining soil contamination from continuing atmospheric or irrigation depositions. Then, for each year of postulated exposure, the code estimates air, surface soil, deep soil, groundwater, and surface water concentrations of each radionuclide in the chain. Human exposures and intakes of each radionuclide are calculated for: 1) pathways of external exposure from finite atmospheric plumes, 2) inhalation, 3) external exposure from contaminated soil, sediments, and water, 4) external exposure from special geometries, and 5) internal exposures from consumption of terrestrial foods, aquatic foods, drinking water, animal products, and inadvertent intake of soil. The intermediate information on annual media concentrations and intake rates are written to data transfer files. Although these may be accessed directly, they are usually used as input to the DOSE module of GENII. GENII is a general purpose computer code used to model dispersion, transport, and long-term exposure effects of specific radionuclides and pathways. Sophisticated codes such as UFOTRI and ETMOD (Environmental Tritium Model) are used exclusively for modeling tritium transport and dosimetry. The UFOTRI and ETMOD codes were not chosen for use in this PEIS because of the lack of information on detailed facility design and on the breakdown of tritium into elemental and tritiated water forms, and because these codes cannot be used for modeling the exposure effects of radionuclides other than tritium. GENII was chosen because it can model both air and surface transport pathways and is not restricted to any radionuclides. DOSE. The DOSE module reads the annual intake and exposure rates defined by the ENV module and converts the data to radiation dose. External dose is calculated with precalculated factors from the EXTDF module or from a data file prepared outside of GENII. Internal dose is calculated with precalculated factors from the INTDF module. EXTDF. The EXTDF module calculates the external dose-rate factors for submersion in an infinite cloud of radioactive materials, immersion in contaminated water, and direct exposure to plane or slab sources of radionuclides. EXTDF was not used. Instead, the dose rate factors listed in External Dose Rate Factors for Calculation of Dose to the Public (DOE/EH-0070) were used for this PEIS. INTDF. Using the Limits for Intakes of Radionuclides by Workers (ICRP Publication 30) model, the INTDF module calculates the internal (inhalation and ingestion) dose conversion factors of radionuclides for specific organs. The factors generated by INTDF were used for the calculations presented in this PEIS. E.2.2.2 Data and Assumptions In order to perform the dose assessments for this PEIS, different types of data must be collected and/or generated. In addition, calculational assumptions have to be made. This section discusses the data collected and/or generated for use in the dose assessment and assumptions made for this PEIS. Meteorological Data. The meteorological data used for all 5 sites were in the form of joint frequency data files. A joint frequency data file is a table listing the fractions of time the wind blows in a certain direction, at a certain speed, and within a certain stability class. The joint frequency data files were based on measurements over a 1-year period at various locations and at different heights at the sites. Average meteorological conditions (averaged over the 1-year period) were used for normal operation. For use in design basis accidents, the 50 percentile option was used. Meteorological data are presented in Health Risk Data (HNUS 1995a). Population Data. Population distributions were based on 1990 Census of Population and Housing data. Projections were determined for the years 2010, 2030, and 2050 for areas within 80 kilometers of the proposed facilities at each candidate site. These years of analysis were selected as representative of the start of operations at tritium supply and recycling facilities, the midplant-life phase, and the end of plant-life phase, respectively. The population was spatially distributed on a circular grid with 16directions and 10 radial distances up to 80 kilometers. The grid was centered on the facility from which the radionuclides were assumed to be released. Population data are presented in Health Risk Data (HNUS 1995a). The site population at the midlife of operation (2030) was assumed to be representative of the population over the 40-year operational period and was used in the impact assessments. Source Term Data. The source terms (quantities of radionuclides released into the environment over a given period) were estimated on the basis of latest conceptual designs of facilities and experience with similar facilities. The source terms used to generate the estimated impacts of normal operation are provided in section E.2.3 for the tritium supply technologies and recycling facilities which could be located at any of the five sites. Source terms for site dependent facilities are presented in sections E.2.3.2 through E.2.3.6. Food Production and Consumption Data. Data from the 1987 Census of Agriculture were used to generate site-specific data for food production. Food production was spatially distributed on the same circular grid as was used for the population distributions. The consumption rates were those used in GENII for the maximum individual and average indi- vidual. People living within the 80 kilometer assessment area were assumed to consume only food grown in that area. Calculational Assumptions. Dose assessments were performed for members of the general public and workers. Dose assessments for members of the public were performed for two different types of receptors considered in this PEIS: a maximally exposed offsite individual and the general population living within 50 miles of the facility. It was assumed that the maximally exposed individual was located at a position on the site boundary that would yield the highest impacts during normal operation of a given alternative. If more than one facility was assumed operating at a site, the dose to this individual from each facility was calculated. The doses were then summed to give the total dose to this individual. A 50-mile population dose was calculated for each operating facility at a site. These doses were then added to give the total population dose at that site. To estimate the radiological impacts from normal operation of reactors and support facilities, additional assumptions and factors were considered in using GENII, as follows: No prior deposition of radionuclides on ground surfaces was assumed. For the maximally exposed offsite individual, the annual exposure time to the plume and to soil contamination was 0.7years (NRC 1977b). For the population, the annual exposure time to the plume and to soil contamination was 0.5 years (NRC 1977b). A semi-infinite/finite plume model was used for air immersion doses. Other pathways evaluated were ground exposure, inhalation, ingestion of food crops and animal products contaminated by either deposition of radioactivity from the air or irrigation, ingestion of fish and other aquatic food raised in contaminated water, swimming and boating in contaminated surface water, and drinking contaminated water. It should be noted that not all pathways were available at every site. For atmospheric releases it was assumed that ground level releases would occur for all tritium supply and recycling facilities. For site-dependent facilities, reported release heights were used and assumed to be the effective stack height. Use of the effective stack height negates plume rise thereby making the resultant doses conservative. The calculated doses were 50-year committed doses from 1 year of intake. Resuspension of particulates was not considered because prior calculations of dust loading in the atmosphere showed that this pathway was negligible compared with others. The exposure, uptake, and usage parameters used in the GENII model are provided in tables E.2.2.2-1 through E.2.2.2-4. Annual average doses to workers for No Action at Idaho National Engineering Laboratory (INEL), Nevada Test Site (NTS), and Pantex Plant (Pantex) were based on measured values received by radiation workers during the 1989 to 1992 time period. The average dose received by a worker at these sites in 2010 was assumed to remain the same as the annual average during the 1989 to 1992 period. The total workforce dose in 2010 was calculated by multiplying the average worker dose by the projected number of workers in 2010. For ORR and SRS, worker dose projections provided by the sites were used. Table E.2.2.2-1.-GENII Exposure Parameters to Plumes and Soil Contamination Maximal Individual General Population External Exposure Inhalation of Plume External Exposure Inhalation of Plume (hours) (hours) Plume Soil Contamination Exposure Time Breathing Rate Plume Soil Exposure Time Breathing Rate (hours) (cc/sec) Contamination (hours) (cc/sec) 6,136 6,136 6,136 270 4,383 4,383 4,383 270 Source: HNUS 1995a. Table E.2.2.2-2.-GENII Usage Parameters for Consumption of Terrestrial Food - Maximum Individual General Population Food Type Growing Yield Holdup Consumption Growing Yield Holdup Consumption Time (kg/m2) Time Rate Time (kg/m2) Time Rate (days) (days) (kg/yr) (days) (days) (kg/yr) Leafy Vegetables 90 1.5 1 30 90 1.5 14 15 Root Vegetables 90 4 5 220 90 4 14 140 Fruit 90 2 5 330 90 2 14 64 Grains/Cereals 90 0.8 180 80 90 0.8 180 72 Source: HNUS 1995a. Table E.2.2.2-3.-GENII Usage Parameters for Consumption of Animal Products - - - Stored Feed Fresh Forage Food Type Consumption Holdup Time Diet Growing Yield Storage Diet Growing Yield Storage Rate (days) Fraction Time (kg/m3) Time Fraction Time (kg/m3) Time (kg/yr) (days) (days) (days) (days) Maximum Individual Beef 80 15 0.25 90 0.8 180 0.75 45 2 100 Poultry 18 1 1 90 0.8 180 - - - - Milk 270 1 0.25 45 2 100 0.75 30 1.5 0 Eggs 30 1 1 90 0.8 180 - - - - General Population Beef 70 34 0.25 90 0.8 180 0.75 45 2 100 Poultry 8.5 34 1 90 0.8 180 - - - - Milk 230 4 0.25 45 2 100 0.75 30 1.5 0 Eggs 20 18 1 90 0.8 180 - - - - Source: HNUS 1995a. Table E.2.2.2-4.-GENII Usage Parameters for Aquatic Activities - Maximum Individual General Population Activity Transit Time Holdup Time Usage Rate Transit Time Holdup Time Usage Rate to Usage Point (days) (per year) to Usage Point (days) (days) (days) Drinking Water 0 0 730 liters 0 0 Site dependent Swimming 0 0 100 hours 0 0 Site dependent Boating 0 0 100 hours 0 0 Site dependent Shoreline 0 0 500 hours 0 0 Site dependent Ingestion of Fish 0 0 40 kg 0 0 Site dependent Ingestion of Mollus 0 0 6.9 kg 0 0 Site dependent Ingestion of Crusta 0 0 6.9 kg 0 0 Site dependent Ingestion of Plants 0 0 6.9 kg 0 0 Site dependent Source: HNUS 1995a. Doses to workers directly associated with tritium supply technologies and recycling facilities were taken from the reports prepared by Fluor Daniel, Inc., and Sandia National Laboratories, New Mexico. To obtain the total workforce dose at a site with a particular tritium supply technology and recycling facilities in operation, the site dose from No Action was added to that from the tritium supply technologies and recycling facilities being evaluated. The average dose to a site worker was then calculated by dividing this dose by the total number of radiation workers at the site. All doses to workers include a component associated with the intake of radioactivity into the body and another component resulting from external exposure to direct radiation. E.2.2.3 Health Effects Calculations Doses calculated by GENII were used to estimate health effects using the risk estimators presented in section E.2.1.2. The incremental cancer fatalities in the general population and in groups of workers from radiation exposure were therefore estimated by multiplying the collective combined effective dose equivalent by 0.0005 and 0.0004 fatal cancers/person-rem, respectively. In this PEIS, the collective combined effective dose equivalent is the sum of the collective committed effective dose equivalent (internal dose) and the collective effective dose equivalent (external dose), section E.2.1.1. Although health risk factors are statistical factors and therefore not strictly applicable to individuals, they have been used in the past to estimate the incremental risk to an individual from exposure to radiation. Therefore, the factors of 0.0005 and 0.0004 per rem of individual committed effective dose equivalent for a member of the public and for a worker, respectively, have also been used in this PEIS to calculate the individual's incremental fatal cancer risk from exposure to radiation. For the public, the health effects expressed in this PEIS are the risk of fatal cancers to the maximally exposed individual and the number of fatal cancers to the 50-mile population from exposure to radioactivity released from any site over the 40-year operational period. For workers, the health effects expressed are the risk to the average worker at a site and the number of fatal cancers to all workers at that site from 40 years of site operations. E.2.3 New Tritium Supply and Recycling Facilities Information This section presents source terms and descriptions of radiological releases to the environment from the normal operation of tritium supply and recycling facilities at Idaho National Engineering Laboratory (INEL), Nevada Test Site (NTS), the Oak Ridge Res- ervation (ORR), Pantex Plant (Pantex), and the Savannah River Site (SRS). In-plant worker doses are also presented. Manufacturing and processing of tritium for use in nuclear weapons is carried out at tritium supply and recycling facilities. There are four different technologies proposed for the production of tritium considered in this PEIS. These are a Heavy-Water Reactor (HWR), a Modular High Temperature Gas-Cooled Reactor (MHTGR), an Advanced Light Water Reactor (ALWR), and Accelerator Production of Tritium (APT). In addition to the reactors, fuel fabrication facilities for the HWR and MHTGR would need to be built to supply fuel to those two types of reactors. Fuel for the ALWR would be procured commercially. For each technology that is proposed for the production of tritium, a tritium extraction facility is needed to remove tritium from the targets. The tritium recycling facility could be collocated with the new tritium supply facility or be located at SRS. This facility processes the tritium received from the tritium supply and from the dismantling of nuclear weapons. For the purposes of this PEIS, the radiological impacts to the public of fuel and target fabrication can be ignored because there are no doses associated with target fabrication and the releases from HWR and MHTGR fuel fabrication are much smaller than from reactor operations. Because the ALWR fuel would be procured commercially, the radiological impacts to workers and the public associated with fuel preparation would not occur at the reactor site and would have already been addressed in environmental statements or reports for the fuel manufacturer. The following subsections present the radiological releases from the facilities associated with tritium supply technologies and recycling facilities. The resulting doses to the public are presented in sections E.2.4 through E.2.8 since they are site dependent. However, because worker doses are dependent on the tritium supply alone, they are included in the following paragraphs. Table E.2.3-1 presents the details of the worker impacts for each reactor/accelerator combination. Table E.2.3-1.-Estimated Annual In-Plant Worker Doses and Resulting Health Effects for Various Tritium Supply Technologies and Recycling Facilities Facility HWR MHTGR ALWR/Large ALWR/Small APT - - - - - Helium-3 SILC Target Target System System Fuel and Target Fabrication Number of badged workers 49 49 0 0 0 0 Person-rem 2.7 0.3 0 0 0 0 Average exposure, mrem 55 5 0 0 0 0 Reactor and Tritium Extraction Number of badged workers 230 180 210 125 258 258 Person-rem 37 28 170 100 38.3 40.8 Average exposure, mrem 161 156 810 800 148 158 Tritium Recycling Number of badged workers 400 400 400 400 400 400 Person-rem 1.6 1.6 1.6 1.6 1.6 1.6 Average exposure, mrem 4 4 4 4 4 4 Totals Number of badged workers 679 629 610 525 658 658 Person-rem 41 30 172 102 40 42 Average exposure, mrem 60 48 281 194 61 64 Annual risk of fatal cancer 2.4x10-5 1.9x10-5 1.1x10-4 7.8x10-5 2.4x10-5 2.6x10-5 (average worker) 40-year plant life, risk of fatal cancers 9.7x10-4 7.6x10-4 4.5x10-3 3.1x10-3 9.7x10-4 1.0x10-3 Additional fatal cancers 0.016 0.012 0.069 0.041 0.016 0.017 annually (all workers) 40-year plant life, additional fatal cancers 0.66 0.48 2.8 1.6 0.64 0.68 Note: SILC - Spallation-induced lithium conversion. Source: DOE 1995d; DOE 1995f; DOE 1995g; SNL 1995a. E.2.3.1 Heavy Water Reactor The Data Report on Heavy Water Reactor (February 1995) describes the HWR considered for this PEIS. Included in the report are the radioactive releases to both the atmosphere and surface water for a "wet" site (ORR and SRS). For a "dry" site (INEL, NTS, and Pantex), it was conservatively assumed that all like radioactive discharges to the surface water at ORR and SRS would be released into the atmosphere. Table E.2.3.1-1 presents these releases. The doses to the worker population for 1 year of operation of the HWR and all its support facilities was calculated to be 41 person-rem. An estimated additional 0.66 fatal cancers could result from 40years of operation. The average worker dose and health effects were calculated to be 60 mrem and a risk of fatal cancer of 9.7x10-4 from 40 years of operation. Table E.2.3.1-1.-Annual Radioactive Releases During Normal Operation from Heavy Water Reactor (curies) Isotope Wet Site Releases Dry Site - Wet Site Releases Dry Site Releases Releases - Liquid Air Air Isotope Liquid Air Air H-3 1.0x103 6.0x103 7.0x103 Sb-124 1.1x10-3 0 1.1x10-3 C-14 0 7.3 7.3 Sb-125 1.1x10-3 0 1.1x10-3 Na-24 0.075 0 0.075 I-129 7.5x10-12 0 7.5x10-12 P-32 5.9x10-6 0 5.9x10-6 I-131 1.8x10-3 8.0x10-3 9.8x10-3 S-35 2.2x10-5 0 2.2x10-5 I-132 1.1x10-3 0 1.1x10-3 Ar-41 0 34 34 I-133 4.2x10-3 0.13 0.13 Cr-51 0.065 0 0.065 I-134 6.7x10-4 0 6.7x10-4 Mn-54 9.3x10-6 0 9.3x10-6 I-135 3.3x10-3 0 3.3x10-3 Fe-59 8.7x10-5 0 8.7x10-5 Xe-133 0 410 410 Co-58 8.0x10-4 0 8.0x10-4 Xe-135m 0 5.5 5.5 Co-60 8.9x10-4 9.6x10-5 9.9x10-4 Xe-135 0 13 13 Zn-65 9.3x10-6 0 9.3x10-6 Cs-134 3.1x10-3 0 3.1x10-3 Kr-85m 0 3.7 3.7 Cs-137 0.061 5.7x10-7 0.061 Kr-85 0 0.8 0.8 Ba-140 3.0x10-3 0 3.0x10-3 Kr-87 0 7.9 7.9 La-140 3.2x10-6 0 3.2x10-6 Kr-88 0 10 10 Ce-141 8.7x10-3 0 8.7x10-3 Sr-89 1.3x10-5 0 1.3x10-5 Ce-143 2.2x10-3 0 2.2x10-3 Sr-90 0.19 4.1x10-4 0.19 Ce-144 6.7x10-3 0 6.7x10-3 Sr-91 2.0x10-3 0 2.0x10-3 Pm-147 6.6x10-7 0 6.6x10-7 Sr-92 7.8x10-4 0 7.8x10-4 U-237 6.3x10-5 0 6.3x10-5 Nb-95 8.6x10-6 0 8.6x10-6 Np-238 3.6x10-5 0 3.6x10-5 Zr-95 0.018 0 0.018 Np-239 0.039 0 0.039 Ru-103 8.8x10-4 0 8.8x10-4 Pu-239 0.012 0 0.012 Source: DOE 1995d. E.2.3.2 Modular High Temperature Gas-Cooled Reactor The Data Report on Modular High Temperature Gas-Cooled Reactor Tritium Supply Plant, February 1995, describes the MHTGR considered for this PEIS. Included in the report are the radioactive releases into the atmosphere. There are no liquid radioactive releases. Table E.2.3.2-1 presents the atmospheric releases. The dose to the worker population for 1 year of operation from the MHTGR and all its support facilities was calculated to be 30 person-rem. An additional 0.48 fatal cancers could result from 40 years of operation. The annual average worker dose was cal- culated to be 48 mrem. A fatal cancer risk of 7.6x10-4 could result to this worker from 40 years of operation. These doses and health effects are given in table E.2.3-1. Table E.2.3.2-1.-Annual Atmospheric Radioactive Releases from Modular High Temperature Gas-Cooled Reactor (curies) Isotope Atmospheric Isotope Atmospheric Isotope Atmospheric Release Release Release H-3 2.1x103 Mo-99 2.0x10-7 I-132 0.31 Kr-85m 1.8 Tc-99m 1.8x10-6 I-133 0.17 Kr-85 4.1x10-3 Ru-103 8.3x10-9 I-134 0.75 Kr-87 3.8 Ru-105 7.9x10-7 I-135 0.27 Kr-88 5.6 Ru-106 7.9x10-11 Xe-133 0.86 Kr-89 2.1 Rh-105 9.4x10-8 Xe-135 1.8 Kr-90 0.86 Ag-110m 8.6x10-13 Cs-134 1.2x10-8 Rb-86 6.4x10-4 Sb-127 4.5x10-9 Cs-136 3.6x10-9 Sr-89 4.5x10-4 Sb-129 5.6x10-6 Cs-137 4.1x10-9 Sr-90 9.0x10-7 Te-127m 5.6x10-5 Ba-140 3.2x10-8 Sr-91 1.2x10-6 Te-127 1.8x10-4 La-140 3.8x10-7 Y-90 1.1x10-8 Te-129m 4.9x10-4 Ce-141 1.7x10-8 Y-91 9.0x10-9 Te-129 0.068 Ce-143 3.6x10-7 Nb-95 1.6x10-8 Te-131m 7.9x10-3 Ce-144 7.5x10-10 Zr-95 8.6x10-9 Te-132 0.018 Pr-143 3.7x10-8 Zr-97 7.5x10-8 I-131 0.026 Nd-147 1.9x10-8 Source: DOE 1995e. E.2.3.3 Advanced Light Water Reactor The Data Report on Advanced Light Water Reactor Tritium Supply Plant, February 1995, evaluated the radiological emissions from the production of tritium using four different advanced light water reactors (two Large ALWRs, Advanced Boiling Water Reactor and CE System 80+ reactor, of about 1,100 to 1,300 MWe and two Small ALWRs, Simplified Boiling Water Reactor and AP600, of about 600MWe). The releases from each are discussed below for both a "dry site" (INEL, NTS, and Pantex) and a "wet site" (ORR and SRS). E.2.3.3.1 Large Advanced Light Water Reactor Radioactive releases into the atmosphere and surface water for a "wet site" (ORR and SRS) and into the atmosphere for a "dry site" (INEL, NTS, and Pantex) from the Advanced Boiling Water Reactor and CE System 80+ reactor are presented in tables E.2.3.3.1-1 and E.2.3.3.1-2, respectively. The dose to the worker population for 1 year of operation from the Large ALWR and all its support facilities was calculated to be 172 person-rem. An additional 2.8 fatal cancers would result from 40 years of operation. The annual average worker dose was calculated to be 281 mrem. A fatal cancer risk of 4.5x10-3 could result to this worker from 40 years of operation. These results and health effects to workers are included in table E.2.3-1. Table E.2.3.3.1-1.-Annual Liquid and Atmospheric Radioactive Releases from Large Advanced Light Water Reactor (Advanced Boiling Water Reactor) (curies) Isotope Wet Site Releases Dry Site - Wet Site Releases Dry Site Releases Releases - Liquid Air Air Isotope Liquid Air Air H-3 7.6x103 8.6x103 1.6x104 Mo-99 8.3x10-4 0.015 0.016 C-14 0.0 9.2 9.2 Tc-99m 8.0x10-4 3.0x10-4 1.1x10-3 Na-24 2.8x10-3 4.1x10-3 6.9x10-3 Ru-103 1.8x10-4 4.9x10-4 6.7x10-4 P-32 1.8x10-4 9.2x10-4 1.1x10-3 Ru-106 1.7x10-4 1.9x10-5 1.9x10-4 Ar-41 0.0 6.8 6.8 Rh-103m 9.0x10-6 1.1x10-4 1.2x10-4 Cr-51 7.7x10-3 0.035 0.043 Ag-110m 3.3x10-4 6.5x10-7 3.3x10-4 Mn-54 2.6x10-3 4.9x10-3 7.5x10-3 Sb-124 3.6x10-4 1.7x10-4 5.3x10-4 Fe-55 5.8x10-3 6.5x10-3 0.012 Te-129m 1.7x10-5 2.2x10-4 2.4x10-4 Mn-56 3.8x10-3 3.5x10-3 7.3x10-3 Te-131m 3.4x10-5 7.6x10-5 1.1x10-4 Fe-59 1.0x10-4 6.5x10-4 7.5x10-4 Te-132 4.0x10-6 1.9x10-5 2.3x10-5 Co-57 7.2x10-5 0.0 7.2x10-5 I-131 3.2x10-3 0.26 0.26 Co-58 9.0x10-5 2.4x10-3 2.5x10-3 I-132 2.6x10-3 2.2 2.2 Co-60 9.1x10-3 0.011 0.02 I-133 0.020 1.7 1.7 Ni-63 1.4x10-4 6.5x10-6 1.5x10-4 I-134 1.7x10-3 3.8 3.8 Cu-64 7.5x10-3 0.010 0.018 I-135 7.5x10-3 2.4 2.4 Zn-65 9.0x10-5 8.1x10-3 8.2x10-3 Xe-131m 0 51 51 Kr-83m 0 8.4x10-4 8.4x10-4 Xe-133m 0 0.086 0.086 Kr-85m 0 21 21 Xe-133 0 2.4x103 2.4x103 Kr-85 0 570 570 Xe-135m 0 410 410 Kr-87 0 25 25 Xe-135 0 460 460 Kr-88 0 38 38 Xe-137 0 510 510 Kr-89 0 240 240 Xe-138 0 430 430 Kr-90 0 3.2x10-4 3.2x10-4 Cs-134 6.1x10-3 1.7x10-4 6.3x10-3 Rb-89 4.4x10-5 4.3x10-5 8.7x10-5 Cs-136 3.2x10-4 8.1x10-5 4.0x10-4 Sr-89 1.1x10-4 5.7x10-3 5.8x10-3 Cs-137 8.9x10-3 4.6x10-4 9.4x10-3 Sr-90 3.5x10-5 6.8x10-5 1.0x10-4 Cs-138 1.9x10-4 1.7x10-4 3.6x10-4 Sr-91 9.0x10-4 1.0x10-3 1.9x10-3 Ba-140 6.8x10-4 0.013 0.014 Sr-92 8.0x10-4 7.8x10-4 1.6x10-3 La-140 1.7x10-4 1.8x10-3 2.0x10-3 Y-90 3.1x10-6 4.6x10-5 4.9x10-5 Ce-141 1.2x10-4 8.6x10-3 8.8x10-3 Y-91 1.1x10-4 2.4x10-4 3.5x10-4 Ce-144 1.9x10-3 1.9x10-5 1.9x10-3 Y-92 6.0x10-4 6.2x10-4 1.2x10-3 Pr-143 1.3x10-6 0 1.3x10-6 Y-93 9.0x10-4 1.1x10-3 2.0x10-3 Pr-144 0 1.9x10-5 1.9x10-5 Nb-95 1.0x10-3 1.7x10-3 2.7x10-3 W-187 9.5x10-5 1.9x10-4 2.8x10-4 Zr-95 8.4x10-4 1.2x10-3 2.0x10-3 Np-239 3.1x10-3 0.012 0.015 Table E.2.3.3.1-2.-Annual Liquid and Atmospheric Radioactive Releases from Large Advanced Light Water Reactor (CE System 80+ Reactor) (curies) Isotope Wet Site Releases Dry Site - Wet Site Releases Dry Site Releases Releases - Liquid Air Air Isotope Liquid Air Air H-3 3.8x103 1.2x104 1.6x104 Ag-110m 2.0x10-3 0 2.0x10-3 C-14 0 7.3 7.3 Sb-124 4.3x10-4 0 4.3x10-4 Na-24 2.2x10-3 0 2.2x10-3 Sb-125 0.0 6.1x10-7 6.1x10-7 P-32 1.8x10-4 0 1.8x10-4 Te-129m 1.0x10-4 0 1.0x10-4 Ar-41 0 34.0 34.0 Te-129 1.2x10-4 0 1.2x10-4 Cr-51 6.4x10-3 3.3x10-5 6.4x10-3 Te-131m 1.6x10-4 0 1.6x10-4 Mn-54 4.8x10-3 2.0x10-5 4.8x10-3 Te-131 3.6x10-5 0 3.6x10-5 Fe-55 8.0x10-3 0.0 8.0x10-3 Te-132 4.1x10-4 0 4.1x10-3 Fe-59 2.4x10-3 8.8x10-6 2.4x10-3 I-131 0.02 0.019 0.039 Co-57 0 2.5x10-6 2.5x10-6 I-132 1.3x10-3 0 1.3x10-3 Co-58 0.011 3.1x10-4 0.011 I-133 0.01 0.053 0.063 Co-60 0.014 9.5x10-5 0.014 I-135 4.3x10-3 0 4.3x10-3 Ni-63 1.7x10-3 0 1.7x10-3 Xe-131m 0 830 830 Zn-65 3.2x10-4 0 3.2x10-4 Xe-133m 0 1 1 Kr-83m 0 8.4x10-4 8.4x10-4 Xe-133 0 660 660 Kr-85m 0 3 3.0 Xe-135m 0 4 4 Kr-85 0 800 800 Xe-135 0 24 24 Kr-87 0 4 4 Xe-138 0 3 3 Kr-88 0 8 8 Cs-139 0 3.1x10-5 3.1x10-5 Sr-89 1.7x10-4 6.9x10-5 2.4x10-4 Cs-134 0.018 3.0x10-5 0.018 Sr-90 2.0x10-5 2.7x10-5 4.7x10-5 Cs-136 8.9x10-4 1.0x10-5 9.0x10-4 Sr-91 2.8x10-5 0 2.8x10-5 Cs-137 0.026 5.2x10-5 0.026 Y-91 9.0x10-5 0 9.0x10-5 Ba-140 7.0x10-3 4.2x10-6 7.0x10-3 Y-91m 1.1x10-5 0 1.1x10-5 La-140 8.4x10-3 0 8.4x10-3 Y-93 1.2x10-4 0 1.2x10-4 Ce-141 3.1x10-4 4.3x10-6 3.2x10-4 Nb-95 2.1x10-3 3.0x10-5 2.1x10-3 Ce-143 3.4x10-4 0 3.4x10-4 Zr-95 1.3x10-3 1.0x10-5 1.3x10-5 Ce-144 6.3x10-3 0 6.3x10-3 Mo-99 1.5x10-3 0.6 1.5x10-3 Pr-143 1.0x10-4 0 1.0x10-4 Tc-99m 1.3x10-3 0 1.3x10-3 Pr-144 2.3x10-3 0 2.3x10-3 Ru-103 4.5x10-3 5.5x10-6 4.5x10-3 W-187 2.1x10-4 0 2.1x10-4 Ru-106 0.065 7.8x10-7 0.065 Np-239 4.4x10-4 0 4.4x10-4 Rh-103m 4.1x10-3 0 4.1x10-3 - - - - E.2.3.3.2 Small Advanced Light Water Reactor Radioactive releases into the atmosphere and surface water for a "wet site" (ORR and SRS) and into the atmosphere for a "dry site" (INEL, NTS, and Pantex) from the Simplified Boiling Water Reactor and AP600 reactor are presented in tables E.2.3.3.2-1 and E.2.3.3.2-2, respectively. The dose to the worker population for 1year of operation from the Small ALWR and all its support facilities was calculated to be 102 person-rem. An additional 1.6 fatal cancers could result from 40 years of operation. The annual average worker dose was calculated to be 194 mrem. A fatal cancer risk of 3.1x10-3 could result to this worker from 40years of operation. These doses and health effects are included in table E.2.3-1. Table E.2.3.3.2-1.-Annual Liquid and Atmospheric Radioactive Releases from Small Advanced Light Water Reactor (Simplified Boiling Water Reactor) (curies) Isotope Wet Site Releases Dry Site - Wet Site Releases Dry Site Releases Releases - Liquid Air Air Isotope Liquid Aira Air H-3 9.9x103 6.3x103 1.6x104 Y-90 3.0x10-6 8.1x10-3 8.1x10-3 C-14 1.6x10-4 5.9 5.9 Y-91 1.1x10-4 0.043 0.043 Ar-41 0 5.4 5.4 Y-92 5.9x10-4 0.068 0.068 Kr-83m 0 4.1x10-4 4.1x10-4 Y-93 8.9x10-4 0.15 0.15 Kr-85m 0 5.7 5.7 Zr-95 8.4x0-4 9.2x10-3 0.01 Kr-85 0 300 300 Nb-95 1.0x10-3 0.02 0.013 Kr-87 0 5.7 5.7 Mo-99 8.4x10-4 0.57 0.57 Kr-88 0 8.6 8.6 Tc-99M 8.1x10-4 0.051 0.052 Kr-89 0 54 54 Ru-103 1.8x10-4 0.022 0.023 Kr-90 0 1.5x10-4 1.5x10-4 Rh-103M 8.9x10-6 0.021 0.021 Xe-131m 0 23 23 Ag-110M 3.2x10-4 8.4x10-7 3.3x10-3 Xe-133m 0 0.027 0.027 Sb-124 3.5x10-4 7.6x10-5 4.3x10-4 Xe-133 0 970 970 Sb-125 0 0 0 Xe-135m 0 89 89 Te-129m 1.7x10-5 0.041 0.041 Xe-135 0 140 140 Te-131M 3.5x10-5 0.013 0.013 Xe-137 0 112 112 Te-131 0 0 0 Xe-138 0 89 89 I-131 3.2x10-3 0.46 0.46 Xe-139 0 1.92x10-4 1.92x10-4 Te-132 4.1x10-6 3.2x10-3 3.2x10-3 Na-24 2.7x10-3 0.092 0.095 I-132 2.6x10-3 1.7 1.7 P-32 1.8x10-4 0.025 0.025 I-133 0.01 1.8 1.8 Cr-51 7.6x10-3 0.89 0.90 I-134 1.7x10-3 2.7 2.7 Mn-54 2.6x10-3 0.014 0.017 Cs-134 6.2x10-3 0.032 0.039 Mn-56 3.8x10-3 0.057 0.061 I-135 7.6x10-3 2.1 2.1 Fe-55 5.7x10-3 0.18 0.18 Cs-136 3.2x10-4 0.014 0.014 Co-57 7.3x10-5 0 7.3x10-5 Cs-137 8.9x10-3 0.084 0.93 Co-58 8.9x10-5 0.032 0.033 Cs-138 1.9x10-4 0.011 0.012 Co-60 9.2x10-3 0.076 0.085 Cs-139 0 1.4x10-5 1.4x10-5 Fe-59 1.00x10-4 5.1x10-3 5.2x10-3 Ba-140 6.8x10-4 0.32 0.33 Ni-63 1.4x10-4 1.8x10-4 3.2x10-4 La-140 1.7x10-4 0.3 0.3 Cu-64 7.6x10-3 0.23 0.24 Ce-141 1.2x10-4 0.035 0.035 Zn-65 8.9x10-5 0.038 0.038 Ce-144 1.9x10-3 3.5x10-3 5.4x10-3 Rb-89 4.3x10-5 2.6x10-3 2.6x10-3 Pr-143 1.3x10-6 0 1.3x10-6 Sr-89 1.1x10-4 0.11 0.11 Pr-144 0 3.5x10-3 3.5x10-3 Sr-90 3.5x10-5 8.1x10-3 8.1x10-3 W-187 9.5x10-5 4.6x10-3 4.7x10-3 Sr-91 8.9x10-4 0.14 0.14 Np-239 3.0x10-3 1.9 1.9 Sr-92 8.1x10-4 0.081 0.082 - - - - Table E.2.3.3.2-2.-Annual Liquid and Atmospheric Radioactive Releases from Small Advanced Light Water Reactor (AP600 Reactor) (curies) Isotope Wet Site Releases Dry Site - Wet Site Releases Dry Site Releases Releases - Liquid Air Air Isotope Liquid Aira Air H-3 1.5x104 1.6x103 1.6x104 Zr-95 1.2x0-3 1.0x10-3 2.2x10-3 C-14 0 7.3 7.3 Nb-95 1.9x10-3 2.5x10-3 4.4x10-3 Ar-41 0 34 34.0 Mo-99 8.1x10-4 0 8.1x10-4 Kr-85m 0 31 31.0 Tc-99M 4.8x10-4 0 4.8x10-4 Kr-85 0 180 180 Ru-103 1.4x10-3 8.0x10-5 1.5x10-3 Kr-87 0 10 10 Rh-103M 1.1x0-3 0 1.1x0-3 Kr-88 0 35 35 Ru106 0.022 7.8x10-5 0.022 Xe-131m 0 1.0x103 1.0x103 Ag-110M 1.4x10-3 0 1.4x10-3 Xe-133m 0 53.0 53 Sb-124 4.3x10-4 0 4.3x10-4 Xe-133 0 2.9x103 2.9x103 Sb-125 0 6.1x10-5 6.1x10-5 Xe-135m 0 4 4.0 Te-129M 3.0x10-5 0 3.0x10-5 Xe-135 0 220 220 Te-129 3.0x10-5 0 3.0x10-5 Xe-138 0 3 3.0 Te-131M 1.4x10-4 0 1.4x10-4 Xe-139 0 1.92x10-4 1.92x10-4 Te-131 3.0x10-5 0 3.0x10-5 Na-24 3.2x10-3 0 3.2x10-3 I-131 0.033 0.066 0.099 P-32 1.8x10-4 0 1.8x10-4 Te-132 2.0x10-4 0 2.0x10-4 Cr-51 5.2x10-3 6.1x10-4 5.8x10-3 I-132 2.3x10-3 0 2.3x10-3 Mn-54 4.4x10-3 4.4x10-4 4.4x10-3 I-133 0.02 0.24 0.26 Fe-55 7.4x10-3 0 7.4x10-3 I-134 9.0x10-5 0 9.0x10-5 Co-57 0 8.2x10-6 8.2x10-6 Cs-134 0.027 0.023 0.029 Co-58 8.6x10-3 0.023 0.032 I-135 0.015 0 0.015 Co-60 0.014 8.7x10-3 0.023 Cs-136 1.5x10-3 8.5x10-5 1.6x10-3 Fe-59 2.2x10-3 7.9x10-5 2.3x10-3 Cs-137 0.037 3.6x10-3 0.041 Ni-63 1.7x10-3 0 1.7x10-3 Ba-140 2.7x10-3 4.2x10-4 3.1x10-3 Zn-65 8.0x10-5 0 8.0x10-5 La-140 2.8x10-3 0 2.8x10-3 Sr-89 1.1x10-4 3.0x10-3 3.1x10-3 Ce-141 2.5x10-4 4.2x10-5 2.9x10-4 Sr-90 2.0x10-5 1.2x10-3 1.2x10-3 Ce-143 2.7x10-4 0 2.7x10-4 Sr-91 5.0x10-5 0 5.0x10-5 Ce-144 4.5x10-3 0 4.5x10-3 Y-91m 3.0x10-5 0 3.0x10-5 Pr-144 5.8x10-4 0 5.8x10-4 Y-91 9.0x10-5 0 9.0x10-5 W-187 2.2x10-4 0 2.2x10-4 Y-93 2.2x10-4 0 2.2x10-4 Np-239 2.5x10-4 0 2.5x10-4 E.2.3.4 Accelerator Production of Tritium Sandia National Laboratories, New Mexico, has evaluated the radiological emissions and Los Alamos National Laboratory has assessed the possible effects of radio frequency power generated, using an accelerator, from the production of tritium. Two separate tritium target designs have been proposed. The first uses targets using the same concept as in the HWR reactor; irradiation of a spallation-induced lithium conversion target. The second uses a different concept; irradiation of helium-3. The accelerator is the same design for either target system. The differences between the two target designs arise in the tritium extraction process discussed below. In addition, a "phased" accelerator with a helium target is being evaluated. Table E.2.3.4-1 presents the atmospheric releases. There are no liquid radioactive releases. Table E.2.3.4-1.-Annual Atmospheric Releases from Accelerator Production of Tritium During Normal Operation (curies) Isotope Full APT Phased APT C-11 11 3.8 N-13 0.49 0.17 O-14 9.8x10-16 3.5x10-16 O-15 6.5x10-8 2.3x10-8 Ar-41 7.1 2.5 Source: SNL 1995a. The radio frequency power generated by the APT accelerator does not produce hazardous exposures of radio frequency non-ionizing radiation to workers, nor does it interfere with radio/television or other sensitive signal equipment, on or offsite. The radio frequency power generators in the APT accelerator system are completely contained in connected, sealed, metallic enclosures. The enclosure provides a continuous metal-sealed envelope which contains the microwave energy. Also, the accelerating cavities and attachments constitute a tightly sealed vacuum system which ensures a much greater shield than is needed for preventing leakage of radio frequency energy. Therefore, the potential for a significant level of microwave energy being radiated to the surroundings of the APT accelerator is negligible. The experience with the Los Alamos Meson Physics Facility at Los Alamos and other DOE accelerators confirms that there is minimal leakage of radio frequency power which poses no hazard to workers or to television/radio or other sensitive signal-processing equipment. The dose to the worker population for 1 year of operation from the APT and all its support facilities with the spallation-induced lithium conversion target was calculated to be 42 person-rem. An additional 0.68 fatal cancers could result from 40 years of operation. The annual average worker dose was calculated to be 64 mrem. A fatal cancer risk of 1.0x10-3 could result to this worker from 40 years of operation. The dose to the worker population for 1 year of operation from the APT and all its support facilities with the helium-3 conversion target was calculated to be 40 person-rem. An additional 0.64 fatal cancers could result from 40 years of operation. The annual average worker dose was calculated to be 61mrem. A fatal cancer risk of 9.7x10-4 could result to this workers from 40 years of operation. A Phased APT is comparable to the APT with the helium-3 target. The dose and health effects associated with all APT options are given in table E.2.3-1. E.2.3.5 Tritium Target Extraction Facility This facility extracts the tritium from the targets after completion of the irradiation in either the reactor or accelerator. Table E.2.3.5-1 presents the atmospheric radiological releases. There are no liquid radioactive releases. The impacts to workers and the public associated with extraction are included in the various reactor/accelerator calculations. Table E.2.3.5-1.-Annual Atmospheric Releases of Tritium from Various Tritium Target Extraction Facilities for Tritium Supply Technologies (curies) Tritium Extraction Facility Atmospheric Contribution Release of Tritium Heavy Water Reactor 5.0x103 Modular High Temperature 5.0x103 Gas-Cooled Reactor Advanced Light Water Reactor 5.0x103 Accelerator - Helium-3 Target 250 Accelerator - Spallation-Induced 5.0x103 Lithium Conversion Target Source: DOE 1995d; DOE 1995e; DOE 1995f; SNL 1995a. E.2.3.6 Tritium Recycling Facility The Data Report on Tritium Recycling Plant (February 1995) describes the facility for the recycling of the tritium. Simply, this facility receives tritium from either the tritium supply or dismantled weapons to purify and prepare the tritium for placement into nuclear weapons. The predicted release of tritium into the atmosphere is estimated to range from 5,000 to 10,000 Ci per year. For this PEIS, the higher value was assumed. There are no liquid radioactive releases. The impacts to workers are included in the various reactor/accelerator calculations. An upgraded tritium recycling facility is being evaluated for operation at SRS. The unconsolidated upgrade involves maintaining many functions in Building 232-H. As a potential mitigation measure a consolidated upgrade is considered that includes the relocation of all tritium processing/handling associated with tritium recycling facilities to other buildings. The predicted tritium releases into the atmosphere from the unconsolidated and consolidated upgrade options are 30,000 Ci per year and 18,000 Ci per year, respectively. E.2.4 Radiological Impacts at Idaho National Engineering Laboratory This section presents the radiological impacts of the tritium supply technologies and recycling facilities at INEL. Section E.2.4.1 presents the radiological releases and resulting impacts from facilities associated with No Action. Section E.2.4.2 presents the radiological releases and resulting impacts from the facilities associated with tritium supply technologies and recycling facilities. For purposes of radiological impact modeling, INEL was divided into five areas which will release radioactivity in the year 2010. All release points in each area were aggregated into a single release point. Table E.2.4-1 presents the characteristics of each of the release points including location, release height, minimum distance, and annual average dispersion to the site boundary in each of the 16 directions. In order to calculate the maximum site boundary dose, the dose from each release point to the maximum receptor associated with each of the other release points has been calculated. For example, the dose resulting from releases from the Test Reactor Area, Argonne National Laboratory-West, Waste Experimental Reduction Facility/Power Burst Facility Area, and the proposed TSS has been determined for the maximum receptor from the Central Facilities Area. Figure E.2.4-1 illustrates the location of each maximum receptor in relation to each release point. The maximum site boundary dose is then determined by the maximum dose of each of the maximum receptors. Table E.2.4-2 presents the direction, distance, and atmospheric dispersion from each release point to each of the maximum receptors. Annual radiological releases were assumed to remain constant during the 40-year operational period. The population and food stuffs distributions centered on each release area are provided in the Health Risk Data report. The joint frequency distribution used for the dose assessment was based on the meteorological measurements for the year 1986 from the GRID III tower at the 10-meter height and is contained in a technical report (Health Risk Data). Figure (Page E-25) Figure E.2.4-1.-Location of Maximum Receptors at Idaho National Engineering Laboratory. Table E.2.4-1.-Release Point Characteristics, Direction, Distance, and Chi/Q at Idaho National Engineering Laboratory Boundary Release Point Argonne National Central Facilities Test Reactor Area Waste Experimental Tritium Supply and Recycling Site Laboratory-West Area Reduction Facility/Power Burst Facility Area Latitude 43o35'41.63" 43o32'4.56" 43o35'8.17" 43o33'3.60" 43o34'42.60" Longitude -112o39'18.71" -112o56'9.95" -112o57'46.79" -112o51'30.95" -11252'5.53" Release Height: 18.1 meters Ground Level 76.2 meters 15.0 meters Ground Level - - - - - - - - - - - Accident Direction Distance Chi/Q Distance Chi/Q Distance Chi/Q Distance Chi/Q Distance Annual 50 95 (m) (sec/m3) (m) (sec/m3) (m) (sec/m3) (m) (sec/m3) (m) Average Percentile Percentile Chi/Q Chi/Q Chi/Q (s/m3) (s/m3) (s/m3) N. 32,636 1.2x10-8 24,783 1.1x10-8 19,099 3.3x10-9 25,459 8.5x10-9 22,328 1.2x10-8 1.2x10-6 5.0x10-6 NNE. 24,647 2.0x10-8 40,102 1.1x10-8 21,737 6.8x10-9 41,136 9.4x10-9 44,888 9.8x10-9 6.1x10-7 2.3x10-6 NE. 19,641 2.3x10-8 45,055 2.3x10-8 42,898 7.3x10-9 39,207 2.4x10-8 37,707 2.9x10-8 6.1x10-7 4.5x10-6 ENE. 16,056 2.2x10-8 39,301 2.0x10-8 41,932 4.6x10-9 32,888 2.1x10-8 34,097 2.5x10-8 6.6x10-7 5.3x10-6 E. 14,466 1.6x10-8 23,843 1.5x10-8 26,375 2.9x10-9 17,582 1.9x10-8 19,375 2.0x10-8 1.2x10-6 7.8x10-6 ESE. 9,003 2.8x10-8 18,763 1.0x10-8 26,410 1.4x10-9 17,858 8.8x10-9 18,696 1.0x10-8 1.5x10-6 6.3x10-6 SE. 5,861 5.0x10-8 11,883 9.4x10-9 19,094 1.0x10-9 14,536 5.9x10-9 18,259 5.2x10-9 1.7x10-6 6.5x10-6 SSE. 5,518 9.2x10-8 10,161 1.7x10-8 15,967 1.6x10-9 11,541 1.1x10-8 14,689 1.0x10-8 2.4x10-6 8.8x10-6 S. 5,572 1.2x10-7 9,887 5.3x10-8 15,538 5.2x10-9 11,539 3.5x10-8 14,635 3.1x10-8 2.2x10-6 8.7x10-6 SSW. 17,066 3.7x10-8 10,022 7.5x10-8 15,754 1.0x10-8 11,938 4.8x10-8 15,029 4.3x10-8 2.0x10-6 8.0x10-6 SW. 19,889 3.1x10-8 11,654 6.4x10-8 18,300 1.2x10-8 13,874 4.2x10-8 17,460 3.7x10-8 4.4x10-7 6.2x10-6 WSW. 28,928 1.7x10-8 16,968 1.5x10-8 18,987 3.3x10-9 20,229 1.0x10-8 25,441 8.9x10-9 9.3x10-7 4.1x10-6 W. 35,299 9.6x10-9 20,727 1.6x10-8 17,013 5.2x10-9 26,938 9.4x10-9 24,305 1.3x10-8 1.1x10-6 5.2x10-6 WNW. 32,525 8.3x10-9 19,193 5.2x10-9 12,186 2.2x10-9 21,125 3.9x10-9 17,918 5.7x10-9 7.7x10-7 6.0x10-6 NW. 27,828 1.0x10-8 17,202 9.7x10-9 11,503 2.9x10-9 20,319 6.4x10-9 17,909 9.2x10-9 1.5x10-6 7.7x10-6 NNW. 31,168 1.3x10-8 17,397 1.0x10-8 12,205 3.4x10-9 23,854 5.7x10-9 20,732 8.2x10-9 1.2x10-6 5.3x10-6 Doses given in this section are associated with 1 year of operation because regulatory standards are given as annual limits. The health effects presented are for the 40-year operational period. Table E.2.4-2.-Direction, Distance, and Meteorological Dispersion to Various Maximum Individual Receptors at Idaho National Engineering Laboratory Site Boundary - Direction Distance Maximum Receptor For: Atmospheric (m) Dispersion Chi/Q (s/m3) Release Point: Central Facilities Area - SSW. 10,021 Central Facilities Area 7.5x10-8 - SW. 15,469 Test Reactor Area 4.3x10-8 - SSE. 10,776 Waste Experimental Reduction 1.5x10-8 Facility/Power Burst Facility - ESE. 22,147 Argonne National Laboratory-West 8.0x10-9 - SSE. 10,329 Tritium Supply Site 1.6x10-8 Release Point: Test Reactor Area - S. 15,549 Central Facilities Area 5.2x10-9 - SW. 18,300 Test Reactor Area 1.2x10-8 - SSE. 16,849 Waste Experimental Reduction 1.5x10-9 Facility/Power Burst Facility - SE. 26,967 Argonne National Laboratory-West 7.4x10-10 - SSE. 16,392 Tritium Supply Site 1.5x10-9 Release Point: WERF/PBF - SW. 14,170 Central Facilities Area 4.1x10-8 - WSW. 21,628 Test Reactor Area 9.2x10-9 - SSW. 11,938 Waste Experimental Reduction 4.8x10-8 Facility/Power Burst Facility - SE. 17,963 Argonne National Laboratory-West 4.4x10-9 - SSW. 12,217 Tritium Supply Site 4.7x10-8 Release Point: ANL-West/EBR-II - WSW. 29,538 Central Facilities Area 1.7x10-8 - WSW. 38,409 Test Reactor Area 1.2x10-8 - SW. 24,904 Waste Experimental Reduction 2.3x10-8 Facility/Power Burst Facility - SSW. 17,066 Argonne National Laboratory-West 3.7x10-8 - SW. 25,872 Tritium Supply Site 2.2x10-8 Release Point: Tritium Supply Site - SSW. 16,430 Central Facilities Area 3.8x10-8 - SW. 22,813 Test Reactor Area 2.6x10-8 - S. 14,873 Waste Experimental Reduction 3.0x10-8 Facility/Power Burst Facility - SE. 20,645 Argonne National Laboratory-West 4.4x10-9 - SSW. 15,029 Tritium Supply Site 4.3x10-8 Note: WERF/PBF - Waste Experimental Reduction Facility/Power Burst Facility; ANL-W/EBR-II Argonne National Laboratory- West/Experimental Breeder Reactor-II. Source: HNUS 1995a. E.2.4.1 No Action Atmospheric Releases. For No Action, four of the areas have radioactive releases into the atmosphere from normal operation. Table E.2.4.1-1 presents the estimated annual atmospheric radioactive releases. Liquid Releases. There are no radioactive liquid releases into the offsite environment associated with No Action. Table E.2.4.1-1.-Annual Atmospheric Radioactive Releases from Normal Operation of No Action at Idaho National Engineering Laboratory (curies) Isotope Argonne National Central Facilities Test Reactor Area Waste Experimental Laboratory-West Area Reduction Facility/ Power Burst Facility H-3 2.3 0 0 0 Na-24 0 0 3.2x10-3 0 Cr-51 0 0 5.1x10-3 0 Mn-54 0 0 2.4x10-6 0 Co-58 0 0 2.4x10-6 0 Co-60 0 1.3x10-6 4.6x10-5 2.1x10-8 Rb-88 0 0 0.5 0 Rb-89 0 0 0.73 0 Sr-90 3.0x10-5 3.7x10-6 4.1x10-4 1.6x10-6 Y-91m 0 0 7.1x10-4 0 Tc-99m 0 0 2.6x10-3 0 Mo-99 0 0 5.3x10-6 0 I-131 2.2x10-4 0 8.3x10-4 4.1x10-6 I-132 9.1x10-4 0 1.2x10-3 0 I-133 3.7x10-4 0 6.3x10-4 0 I-134 0 0 1.1x10-3 0 Cs-134 0 0 9.6x10-7 0 Cs-137 0 6.3x10-8 3.2x10-6 1.8x10-7 Cs-138 0 0 0.71 0 Ba-139 0 0 0.051 0 Ba-140 2.3x10-3 0 9.4x10-6 0 La-140 0 0 1.4x10-5 0 Hg-203 0 0 3.6x10-4 0 Os-191 0 0 6.6x10-4 0 Pu-239 8.0x10-6 9.6x10-7 7.4x10-7 1.3x10-7 Ar-41 57 0 3.2x103 0 Kr-85 0.18 0 0 0 Kr-85m 21 0 22 0 Kr-87 24 0 68 0 Kr-88 24 0 66 0 Xe-131m 0.1 0 0 0 Xe-133 450 0 25 0 Xe-133m 1.8x10-3 0 0 0 Xe-135 120 0 79 0 Xe-135m 5.5 0 40 0 Xe-138 15 0 200 0 Tables E.2.4.1-2 and E.2.4.1-3 include the radiological impacts to the maximally exposed individual and offsite population within 50 miles, respectively. The maximally exposed individual would receive an annual dose of 6.0x10-3 mrem. An estimated fatal cancer risk of 1.2x10-7 would result from 40 years of operation. The population within 50 miles would receive a dose of 0.037 person-rem in year 2030 (mid-life of operation). An estimated 7.4x10-4 fatal cancers could result from 40 years of operation. Table E.2.4.1-2.-Doses and Resulting Health Effect to the Maximally Exposed Individual Resulting from Normal Operation at Idaho National Engineering Laboratory - Dose by Pathway (mrem per year) - - - Alternative Inhalation Ingestion Plume Ground Shine Committed Percent of Estimated Immersion Effective Dose Background 40-Year Fatal Equivalent Cancer Risk (mrem per year) No Action 3.1x10-5 1.7x10-5 5.9x10-3 9.5x10-6 6.0x10-3 1.7x10-3 1.2x10-7 No Action Contribution to Tritium Supply Site 2.1x10-5 1.7x10-5 1.2x10-3 2.6x10-6 1.2x10-3 3.5x10-4 2.5x10-8 Maximum Receptor Heavy Water Reactor 0.039 0.14 4.2x10-4 1.8x10-4 0.18 0.052 3.7x10-6 Modular High Temperature Gas-Cooled 5.2x10-3 0.071 1.0x10-4 5.1x10-5 0.077 0.022 1.5x10-6 Reactor Advanced Boiling Water Reactor 0.016 0.24 3.2x10-3 7.0x10-4 0.25 0.072 5.1x10-6 CE System 80+ 0.016 0.23 4.9x10-4 3.2x10-4 0.24 0.069 4.9x10-6 Simplified Boiling Water Reactor 0.016 0.24 8.9x10-4 1.7x10-3 0.25 0.072 5.1x10-6 AP600 0.016 0.23 1.5x10-3 4.9x10-4 0.24 0.069 4.9x10-6 Full Accelerator Production of Tritium with 1.8x10-4 2.5x10-3 1.5x10-5 6.6x10-10 2.7x10-3 7.7x10-4 5.4x10-8 helium-3 Full Accelerator Production of Tritium with 3.6x10-3 0.049 1.5x10-5 6.6x10-10 0.053 0.015 1.1x10-6 spallation-induced lithium conversion Phased Accelerator Production of Tritium 1.8x10-4 2.5x10-3 7.6x10-6 3.4x10-10 2.7x10-3 7.7x10-4 5.4x10-8 Tritium Recycling 7.3x10-3 0.099 6.9x10-12 0 0.11 0.031 2.2x10-6 Worker Doses. Based on measured values during the time period of 1989 to 1992 (Twenty-Second Annual Report Radiation Exposures for DOE and DOE Contractor Employees - 1989 (DOE/EH-0286P)) and subsequent yearly data reports), the annual average dose to a badged worker at INEL was calculated to be 30 mrem. It is projected that in the year 2010 and beyond, there would be 7,337 badged workers involved in No Action activities at INEL (INEL 1993a:5). The annual average dose to these workers is assumed to remain at 30 mrem; the annual total dose among all these workers would then equal 220 person-rem. From 40 years of operation, an estimated fatal cancer risk of 4.8x10-4 would result to the average worker and 3.5 fatal cancers could result among all workers. Table E.2.4.1-3.-Doses and Resulting Health Effect to the Population Within 50 Miles Resulting from Normal Operation at Idaho National Engineering Laboratory Alternative Dose by Pathway in 2030 (person-rem) - - - - Inhalation Ingestion Plume Ground Shine Committed Percent of Estimated Immersion Effective Dose Background 40-Year Fatal Equivalent in Cancers 2030 (person-rem) No Action 4.1x10-4 8.0x10-3 0.029 4.6x10-5 0.037 7.0x10-5 7.4x10-4 Heavy Water Reactor 0.33 31 1.8x10-3 1.5x10-3 31 0.058 0.62 Modular High Temperature 0.044 15 4.1x10-4 3.3x10-4 15 0.027 0.29 Gas-Cooled Reactor Advanced Boiling Water Reactor 0.13 51 0.011 4.8x10-3 51 0.096 1 CE System 80+ 0.13 49 2.5x10-3 2.7x10-3 49 0.092 0.98 Simplified Boiling Water Reactor 0.13 49 3.6x10-3 0.014 49 0.092 0.98 AP600 0.13 49 8.9x10-3 4.1x10-3 49 0.092 0.98 Full Accelerator Production of Tritium 1.5x10-3 0.53 4.7x10-5 3.7x10-11 0.53 1.0x10-3 0.011 with helium-3 Full Accelerator Production of Tritium with 0.031 10 4.7x10-5 3.7x10-11 10 0.019 0.2 spallation-induced lithium conversion Phased Accelerator Production of Tritium 1.5x10-3 0.53 2.4x10-5 1.9x10-11 0.53 1.0x10-3 0.011 Tritium Recycling 0.061 21 5.8x10-11 0 22 0.041 0.44 E.2.4.2 Tritium Supply Technologies and Recycling For the tritium supply technologies and recycling facilities, the impacts from the No Action facilities need to be added to the impacts from the tritium supply technologies and recycling facilities to determine the impacts from total site operation. For example, to determine the radiological impact for the addition of an HWR at INEL, the doses from No Action facilities would be summed with the HWR doses (which includes tritium target extraction) and the tritium recycling doses. Estimated annual atmospheric radioactive releases for the tritium supply technologies and recycling facilities are given in section E.2.3. Tables E.2.4.1-2 and E.2.4.1-3 present the radiological impacts by tritium supply technology and recycling facility. There are no radioactive liquid releases to the offsite environment associated with this alternative. The annual doses from total site operation associated with the difference tritium supplies ranged from 0.11to 0.36 mrem to the maximally exposed individual and from 23 to 73 person-rem to the 50-mile population in the year 2030. The health effects from 40years of operation are included in both tables. E.2.5 Radiological Impacts at Nevada Test Site This section presents the radiological impacts of the tritium supply technologies and recycling facilities at NTS. Section E.2.5.1 presents the radiological releases and resulting impacts from facilities associated with No Action. Section E.2.5.2 presents the radiological releases and resulting impacts from the facilities associated with tritium supply technologies and recycling facilities. For purposes of radiological impact modeling, NTS was divided into seven areas which will release radioactivity in the year 2010. All release points in each area were aggregated into a single release point. Table E.2.5-1 presents the characteristics of each of the release points including location, release height, minimum distance, and annual average dispersion to the site boundary in each of the 16 directions. In order to calculate the maximum site boundary dose, the dose from each release point to the maximum receptor associated with each of the other release points has been calculated. For example, the dose resulting from releases for Areas 5, 6, 12, 19, 23, and the proposed TSS has been determined from the maximum receptor for Area 3. Figure E.2.5-1 illustrates the location of each maximum receptor in relation to each release point. The maximum site boundary dose is then determined by the maximum dose of each of the maximum receptors. Table E.2.5-2 presents the direction, distance, and atmospheric dispersion from each release point to each of the maximum receptors. Annual radiological releases were assumed to remain constant during the 40-year operational period. The population and food stuffs distributions centered on each release area are provided in the Health Risk Data report. The joint frequency distribution used for the dose assessment was based on the meteorological measurements for the year 1990 from Desert Rock at the 10 meter height and is contained in a technical report (Health Risk Data). Figure (Page E-33) Figure E.2.5-1.-Location of Maximum Receptors at Nevada Test Site. Table E.2.5-1.-Release Point Characteristics, Direction, Distance, and Chi/Q at Nevada Test Site Boundary [Page 1 of 2] Release Point Area 3 Area 5 Area 6 Area 12 Latitude 37o1'36.12" 36o46'47.64" 36o56'816" 37o11'48.48" Longitude -116o0'39.25" -116o0'10.08" -116o4'623" -116o11'10.69" Release Height: Ground Level Ground Level Ground Level Ground Level Direction Distance Chi/Q Distance Chi/Q Distance Chi/Q Distance Chi/Q (m) (s/m3) (m) (s/m3) (m) (s/m3) (m) (s/m3) N. 24,166 8.7x10-9 32,405 5.9x10-9 34,911 5.4x10-9 5,451 6.5x10-8 NNE. 13,134 2.1x10-8 11,810 2.5x10-8 22,182 1.1x10-8 5,569 6.9x10-8 NE. 8,858 4.8x10-8 7,952 5.6x10-8 14,928 2.4x10-8 6,606 7.3x10-8 ENE. 7,522 5.3x10-8 6,753 6.1x10-8 12,727 2.6x10-8 9,858 3.7x10-8 E. 7,392 5.2x10-8 6,632 6.0x10-8 12,500 2.5x10-8 22,944 1.1x10-8 ESE. 7,562 4.2x10-8 6,783 4.9x10-8 12,775 2.0x10-8 23,454 9.1x10-9 SE. 8,969 3.1x10-8 8,078 3.5x10-8 15,148 1.5x10-8 27,813 6.7x10-9 SSE. 13,549 1.9x10-8 12,243 2.2x10-8 22,834 9.7x10-9 42,052 4.4x10-9 S. 39,154 6.0x10-9 12,279 2.8x10-8 30,398 8.4x10-9 58,365 3.6x10-9 SSW. 41,092 6.2x10-9 17,496 1.9x10-8 30,294 9.2x10-9 40,688 6.3x10-9 SW. 46,379 6.6x10-9 17,352 2.4x10-8 36,377 9.0x10-9 27,188 1.3x10-8 WSW. 39,124 5.7x10-9 22,735 1.2x10-8 34,550 6.8x10-9 23,618 1.1x10-8 W. 38,373 8.3x10-9 39,856 7.9x10-9 33,894 9.8x10-9 24,146 1.5x10-8 WNW. 39,139 8.4x10-9 40,669 7.9x10-9 34,562 9.8x10-9 27,927 1.3x10-8 NW. 46,438 3.8x10-9 48,027 3.7x10-9 40,791 4.5x10-9 23,860 9.0x10-9 NNW. 24,654 3.8x10-9 53,210 1.4x10-9 35,691 2.4x10-9 16,125 6.7x10-9 Release Pointa Area 19 Area 23 Tritium Supply and Recycling Site Latitude 37o18'51.84" 36o39'13.33" 36o53'42.36" Longitude -116o19'20.29" -115o59'52.45" -116o2'54.96" Release Height: Ground Level Ground Level 10 meters - - - - - - - Accident Direction Distance Chi/Q Distance Chi/Q Distance Annual 50 95 (m) (s/m3) (m) (s/m3) (m) Average Percentile Percentile Chi/Q Chi/Q Chi/Q (s/m3) (s/m3) (s/m3) N. 6,883 4.7x10-8 30,320 6.5x10-9 37,260 5.0x10-9 2.5x10-8 3.9x10-6 NNE. 7,027 5.0x10-8 11,023 2.7x10-8 14,516 1.9x10-8 4.5x10-8 9.0x10-6 NE. 7,397 6.2x10-8 1,993 4.5x10-7 9,785 4.2x10-8 6.9x10-8 1.9x10-5 ENE. 8,018 4.8x10-8 1,652 5.2x10-7 8,319 4.6x10-8 1.1x10-7 2.9x10-5 E. 10,447 3.2x10-8 1,607 5.0x10-7 8,175 4.5x10-8 1.7x10-7 4.3x10-5 ESE. 12,243 2.2x10-8 1,629 4.1x10-7 8,359 3.7x10-8 5.7x10-7 4.2x10-5 SE. 42,415 3.9x10-9 1,899 3.0x10-7 9,922 2.7x10-8 1.0x10-6 3.4x10-5 SSE. 64,080 2.5x10-9 2,827 1.8x10-7 14,923 1.7x10-8 4.7x10-7 2.0x10-5 S. 53,585 4.0x10-9 6,501 6.8x10-8 27,173 9.7x10-9 1.0x10-7 7.2x10-6 SSW. 20,987 1.5x10-8 7,059 6.5x10-8 29,216 9.6x10-9 3.1x10-8 3.9x10-6 SW. 18,429 2.2x10-8 7,785 7.0x10-8 33,046 1.0x10-8 3.4x10-8 3.1x10-6 WSW. 18,295 1.6x10-8 7,558 5.3x10-8 38,968 5.8x10-9 1.1x10-7 2.9x10-6 W. 12,833 3.6x10-8 9,118 5.7x10-8 38,242 8.4x10-9 9.9x10-8 4.3x10-6 WNW. 8,875 6.0x10-8 41,220 7.8x10-9 39,002 8.4x10-9 6.3x10-8 4.2x10-6 NW. 8,117 3.9x10-8 48,728 3.6x10-9 46,033 3.8x10-9 3.8x10-8 3.7x10-6 NNW. 7,018 2.1x10-8 67,499 1.0x10-9 38,079 2.2x10-9 4.5x10-8 5.7x10-6 Doses given in this section are associated with 1 year of operation because regulatory standards are given as annual limits. The health effects presented are for the 40-year operational period. Table E.2.5-2.-Direction, Distance, and Meteorological Dispersion to Various Maximum Individual Receptors at Nevada Test Site Boundary [Page 1 of 2] Direction Distance Maximum Receptor For: Atmospheric Dispersion (m) Chi/Q (sec/m3) Release Point: Area 3 ENE. 7,522 Area 3 5.3x10-8 SSE. 27,103 Area 5 7.8x10-9 SE. 10,672 Area 6 2.4x10-8 NNW. 27,004 Area 12 3.4x10-9 NNW. 44,680 Area 19 1.8x10-9 SE. 13,243 Tritium Supply Site 1.8x10-8 S. 41,158 Area 23 5.6x10-9 Release Point: Area 5 NNE. 29,543 Area 3 7.5x10-9 ENE. 6,752 Area 5 6.1x10-8 NNE. 20,762 Area 6 1.2x10-8 NNW. 53,166 Area 12 1.4x10-9 NNW. 69,742 Area 19 1.0x10-9 NNE. 17,680 Tritium Supply Site 1.4x10-8 S. 13,841 Area 23 2.4x10-8 Release Point: Area 6 NE. 17,000 Area 3 2.0x10-8 SE. 20,273 Area 5 1.0x10-8 ENE. 12,727 Area 6 2.6x10-8 NNW. 35,041 Area 12 2.4x10-9 NNW. 51,517 Area 19 1.5x10-9 E. 12,532 Tritium Supply Site 2.5x10-8 SSE. 31,953 Area 23 6.3x10-9 Release Point: Area 12 SE. 28,863 Area 3 6.4x10-9 SSE. 50,480 Area 5 3.5x10-9 SE. 35,133 Area 6 4.9x10-9 NE. 6,605 Area 12 7.3x10-8 NNW. 20,708 Area 19 4.8x10-9 SE. 37,683 Tritium Supply Site 4.5x10-9 SSE. 62,700 Area 23 2.6x10-9 Direction Distance Maximum Receptor For: Atmospheric Dispersion (m) Chi/Q (s/m3) Release Point: Area 19 SE. 46,455 Area 3 3.4x10-9 SSE. 67,757 Area 5 2.4x10-9 SE. 52,890 Area 6 2.9x10-9 ESE. 17,575 Area 12 1.3x10-8 NE. 7,396 Area 19 6.2x10-8 SE. 55,398 Tritium Supply Site 2.8x10-9 SSE. 79,095 Area 23 1.9x10-9 Release Point: Tritium Supply Site NNE. 16,170 Area 3 1.6x10-8 SSE. 15,807 Area 5 1.6x10-8 ENE. 9,495 Area 6 3.8x10-8 NNW. 38,441 Area 12 2.2x10-9 NNW. 55,421 Area 19 1.4x10-9 ENE. 8,319 Tritium Supply Site 4.6x10-8 S. 28,747 Area 23 9.0x10-9 Release Point: Area 23 N. 43,225 Area 3 4.1x10-9 NNE. 16,488 Area 5 1.6x10-8 N. 34,229 Area 6 5.5x10-9 NNW. 66,933 Area 12 1.1x10-9 NNW. 83,144 Area 19 8.1x10-10 N. 31,002 Tritium Supply Site 6.3x10-9 ENE. 1,651 Area 23 5.2x10-7 Source: HNUS 1995a. E.2.5.1 No Action Atmospheric Releases. For No Action, six of the areas have radioactive releases into the atmosphere from normal operation. Table E.2.5.1-1 presents the estimated annual atmospheric radioactive releases. Liquid Releases. There are no radioactive liquid releases into the offsite environment associated with No Action. Tables E.2.5.1-2 and E.2.5.1-3 respectively, include the radiological impacts to the maximally exposed member of the public and offsite population within 50 miles. The maximally exposed individual would receive an annual dose of 0.040 mrem. An estimated fatal cancer risk of 8.1x10-7 would result from 40years of operation. The population within 50miles would receive a dose of 8.2x10-3 person-rem in year 2030 (mid-life of operation). An estimated 1.6x10-4 fatal cancers could result from 40years of operation. Workers Doses. Based on measured values during the time period of 1989 to 1992 (Twenty-Second Annual Report Radiation Exposures for DOE and DOE Contract Employees - 1989 (DOE/EH-0286P)) and subsequent yearly dose reports), the annual average dose to a badged worker at NTS was calculated to be 5 mrem. It is projected that in the year 2010 and beyond, there would be 619 badged workers involved in No Action activities at NTS (NTS 1993a:4). The annual average dose to these workers was assumed to remain at 5 mrem; the annual total dose among all these workers would then equal 3 person-rem. From 40 years of operation, an estimated fatal cancer risk of 7.8x10-5 would result to the average worker and 0.048 fatal cancer could result among all workers. Table E.2.5.1-1.-Estimated Annual Atmospheric Radioactive Releases from Normal Operation of No Action at Nevada Test Site (curies) Isotope Area 3 Area 5 Area 6 Area 12 Area 19 Area 23 H-3 0 0.6 4.8x10-3 2.2x103 0 5.1x10-4 Ar-37 0 0 0 0 0 0 Ar-39 0 0 0 0 0 0 Kr-85 0 0 0 0 280 0 Xe-127 0 0 0 0 0 0 Xe-129 0 0 0 0 0 0 Xe-131 0 0 0 0 0 0 Xe-133 0 0 0 0 0 0.04 I-131 0 0 1.3x10-5 0 0 5.8x10-5 Pu-239 2.5x10-3 0 0 0 0 0 U-234 0 0 0 0 0 0 U-235 0 0 0 0 0 0 U-238 0 0 0 0 0 0 Note: For reconfiguration, Device Assembly Facility upgrade assumed to be equivalent to the assembly, disassembly and high explosion upgrade at Pantex. Source: NTS 1993a:4; table E.2.7-2. Table E.2.5.1-2.-Doses and Resulting Health Effect to the Maximally Exposed Individual Resulting from Normal Operation at Nevada Test Site Alternative Dose by Pathway (mrem per year) - Percent of Estimated Background 40-Year Fatal Cancer Risk - Inhalation Ingestion Plume Ground Shine Committed - - Immersion Effective Dose Equivalent (mrem per year) No Action 3.2x10-3 0.038 1.5x10-6 1.1.x10-9 0.04 0.013 8.1x10-7 No Action Contribution to 2.8x10-3 2.3x10-3 3.0x10-7 8.0x10-9 5.1x10-3 1.6x10-3 1.0x10-7 Tritium Supply Site Maximum Receptor Heavy Water Reactor 0.039 0.14 4.8x10-4 1.8x10-4 0.19 0.059 3.7x10-6 Modular High Temperature 5.5x10-3 0.075 1.2x10-4 5.6x10-5 0.081 0.026 1.6x10-6 Gas-Cooled Reactor Advanced Boiling Water Reactor 0.017 0.25 5.1x10-3 8.0x10-4 0.28 0.087 5.5x10-6 CE System 80+ 0.017 0.24 6.0x10-4 3.4x10-4 0.27 0.084 5.3x10-6 Simplified Boiling Water Reactor 0.017 0.25 1.3x10-3 1.9x10-3 0.28 0.087 5.5x10-6 AP600 0.017 0.24 1.7x10-3 5.2x10-4 0.27 0.084 5.3x10-6 Full Accelerator Production of 2.0x10-4 2.7x10-3 1.4x10-5 7.0x10-9 2.9x10-3 9.2x10-4 5.8x10-8 Tritium with helium-3 Full Accelerator Production of 3.9x10-3 0.053 1.4x10-5 7.0x10-9 0.057 0.018 1.1x10-6 Tritium with spallation- induced lithium conversion Phased Accelerator Production 2.0x10-4 2.7x10-3 7.2x10-6 3.6x10-9 2.9x10-3 9.2x10-4 5.8x10-8 of Tritium Tritium Recycling 7.9x10-3 0.11 7.2x10-12 0 0.12 0.038 2.4x10-6 Table E.2.5.1-3.-Doses and Resulting Health Effect to the Population Within 50 Miles Resulting from Normal Operation at Nevada Test Site Alternative Dose by Pathway in 2030 (person-rem) - Percent of Estimated Background 40-Year Fatal Cancers - Inhalation Ingestion Plume Ground Shine Committed - - Immersion Effective Dose Equivalent in 2030 (person-rem) No Action 4.3x10-3 3.9x10-3 1.1x10-5 5.5x10-9 8.2x10-3 1.4x10-4 1.6x10-4 Heavy Water Reactor 0.035 0.083 2.4x10-4 1.7x10-4 0.12 2.0x10-3 2.4x10-3 Modular High Temperature 4.9x10-3 0.045 5.8x10-5 3.9x10-5 0.051 8.7x10-4 1.0x10-3 Gas-Cooled Reactor Advanced Boiling Water Reactor 0.014 0.15 1.4x10-3 5.5x10-4 0.16 2.8x10-3 3.3x10-3 CE System 80+ 0.014 0.14 3.2x10-4 3.0x10-4 0.16 2.6x10-3 3.1x10-3 Simplified Boiling Water Reactor 0.014 0.16 4.6x10-4 1.6x10-3 0.17 3.0x10-3 3.5x10-3 AP600 0.014 0.15 1.1x10-3 4.5x10-4 0.16 2.8x10-3 3.3x10-3 Full Accelerator Production of 3.4x10-3 0.031 7.1x10-6 6.7x10-13 0.035 6.0x10-4 7.0x10-4 Tritium with SILC Full Accelerator Production of 1.7x10-4 1.6x10-3 7.6x10-6 6.7x10-13 1.8x10-3 3.1x10-5 3.6x10-5 Tritium with helium-3 Phased Accelerator Production of 1.7x10-4 1.6x10-3 3.6x10-6 3.4x10-13 1.8x10-3 3.1x10-5 3.6x10-5 Tritium Tritium Recycling 6.9x10-3 0.063 6.4x10-12 0 0.07 1.2x10-3 1.4x10-3 E.2.5.2 Tritium Supply Technologies and Recycling For the tritium supply technologies and recycling facilities, the impacts from the No Action facilities need to be added to the impacts from the tritium supply technologies and recycling facilities to determine the impacts from total site operation. For example, to determine the radiological impact for the addition of an HWR at NTS, the doses from No Action facilities would be summed with the HWR doses (which includes tritium target extraction) and the tritium recycling doses. Estimated annual atmospheric radioactive releases for the tritium supply technologies and recycling facilities are given in section E.2.3. Tables E.2.5.1-2 and E.2.5.1-3 present the radiological impacts by tritium supply technology and recycling facility. There are no radioactive liquid releases to the offsite environment associated with this alternative. The annual doses from total site operation associated with the different tritium supplies ranged from 0.12to 0.40 mrem to the maximally exposed individual and from 0.1 to 0.24 person-rem to the 50-mile population in the year 2030. The health effects from 40 years of operation are included in both tables. E.2.6 Radiological Impacts at Oak Ridge Reservation This section presents the radiological impacts of the tritium supply technologies and recycling facilities at ORR. Section E.2.6.1 presents the radiological releases and resulting impacts from facilities associated with No Action. Section E.2.6.2 presents the radiological releases and resulting impacts from the facilities associated with the tritium supply technologies and recycling facilities. For purposes of radiological impact modeling, ORR was divided into six areas. All potential release points in each area were aggregated into a single release point. Table E.2.6-1 presents the characteristics of each of the release points including location, release height, minimum distance, and annual average dispersion to the site boundary in each of 16directions. In order to calculate the maximum site boundary dose, the dose from each release point to the maximum receptor associated with each of the other release points has been calculated. For example, the dose resulting from releases from the Oak Ridge National Laboratory, Y-12 Plant, Advanced Neutron Source Reactor, High Flux Isotope Reactor areas, and the proposed TSS has been determined for the maximum receptor from the K-25 incinerator. Figure E.2.6-1 illustrates the location of each maximum receptor in relation to each release point. The maximum site boundary dose is then determined by the maximum dose of each of the maximum receptors. Table E.2.6-2 presents the distance, direction, and atmospheric dispersion from each release point to each of the maximum receptors. Annual radiological releases were assumed to remain constant during the 40-year operation period. Figure (Page E-43) Figure E.2.6-1.-Location of Maximum Receptors at Oak Ridge Reservation. Table E.2.6-1.-Release Point Characteristics, Direction, Distance, and Chi/Q at Oak Ridge Reservation Boundary [Page 1 of 2] Release Pointa K-25 Incinerator ORNL (3039 Stack) Y-12 Plant REDC Latitude 35o56'15.36" 35o55'39.01" 35o59'08.52" 35o55'08.04" Longitude -84o22'54.83" -84o18'55.44" -84o15'38.51" -84o18'12.95" Release Height: 30.5 meters 76.2 meters 20.0 meters 76.2 meters Direction Distance Chi/Q Distance Chi/Q Distance Chi/Q Distance Chi/Q (meter) (sec/m3) (meter) (sec/m3) (meter) (sec/m3) (meter) (sec/m3) N. 3,042 3.5x10-7 4,202 1.0x10-7 683 1.1x10-6 5,838 4.5x10-8 NNE. 3,933 4.1x10-7 5,845 1.8x10-7 888 1.5x10-6 7,329 1.1x10-7 NE. 4,361 5.3x10-7 8,512 1.9x107 1,628 8.1x10-7 3,381 7.1x10-7 ENE. 4,634 3.4x10-7 3,933 6.1x10-7 2,336 4.6x10-7 3,440 7.5x10-7 E. 9,767 7.1x10-8 4,358 3.5x10-7 2,964 3.7x10-7 3,122 5.6x10-7 ESE. 9,644 6.0x10-8 4,390 1.9x10-7 2,283 1.4x10-7 3,046 2.2x10-7 SE. 4,924 1.9x10-7 4,029 1.5x10-7 2,329 8.1x10-8 2,595 1.6x10-7 SSE. 2,307 5.3x10-7 4,358 1.5x10-7 3,728 8.2x10-8 2,913 8.6x10-8 S. 2,418 4.1x10-7 4,299 2.5x10-7 4,682 6.1x10-8 3,350 8.5x10-8 SSW. 3,303 2.9x10-7 3,752 4.0x10-7 9,615 4.2x10-8 3,449 1.6x10-7 SW. 3,897 4.2x10-7 3,755 2.8x10-7 11,872 1.1x10-7 3,626 5.0x10-7 WSW. 2,894 6.4x10-7 5,054 2.6x10-7 3,442 5.3x10-7 3,895 6.2x10-7 W. 3,619 1.7x10-7 8,677 5.8x10-8 1,074 7.7x10-7 5,945 1.0x10-7 WNW. 2,782 1.6x10-7 7,260 2.7x10-8 806 5.5x10-7 8,636 2.2x10-8 NW. 2,356 2.2x10-7 4,464 4.6x10-8 686 6.5x10-7 5,267 3.2x10-8 NNW. 1,857 3.0x10-7 3,900 7.8x10-8 620 8.1x10-7 5,262 3.5x10-8 Release Point ANSR Tritium Supply and Recycling Site Latitude 35o55'32.52" 35o55'59.17" Longitude -84o17'31.20" -84o20'55.68" Release Height: 10.0 meters Ground Level - - - - - Accident Direction Distance Chi/Q Distance Annual 50 Percentile 95 Percentile (meter) (sec/m3) (meter) Chi/Q Chi/Q Chi/Q (sec/m3) (s/m3) (s/m3) N. 5,859 4.5x10-8 3,199 2.2x10-7 1.8x10-5 1.3x10-4 NNE. 6,940 1.2x10-7 2,993 5.8x10-7 1.9x10-5 1.7x10-4 NE. 2,091 1.5x10-6 4,638 6.2x10-7 1.6x10-5 8.4x10-5 ENE. 2,124 1.6x10-6 9,495 2.9x10-7 5.8x10-6 3.1x10-5 E. 2,815 6.6x10-7 6,807 1.5x10-7 6.4x10-6 3.6x10-5 ESE. 2,483 3.0x10-7 6,781 1.2x10-7 5.8x10-6 2.6x10-5 SE. 2,465 1.8x10-7 5,908 6.9x10-8 7.6x10-6 2.9x10-5 SSE. 2,679 9.9x10-8 3,577 6.0x10-8 9.0x10-6 3.5x10-5 S. 3,172 9.3x10-8 3,418 8.7x10-8 1.1x10-5 5.2x10-5 SSW. 4,311 1.1x10-7 3,098 4.5x10-7 1.9x10-5 1.1x10-4 SW. 4,903 3.2x10-7 2,903 1.1x10-6 2.0x10-5 1.5x10-4 WSW. 5,215 4.0x10-7 4,900 2.1x10-7 9.7x10-6 8.0x10-5 W. 7,350 7.5x10-8 5,700 5.6x10-8 6.8x10-6 5.7x10-5 WNW. 8,891 2.1x10-8 4,294 4.7x10-8 6.2x10-6 5.7x10-5 NW. 5,174 3.3x10-8 4,787 3.9x10-8 5.1x10-6 5.4x10-5 NNW. 5,181 3.6x10-8 4,770 4.7x10-8 3.5x10-6 4.1x10-5 Descriptions of population, food stuffs distributions, and aquatic foods for each release area are provided in the Health Risk Data report. The joint frequency distributions used for the dose assessment were based on 1990 meteorological measure- ments from five meteorological towers (Tower 1 for K-25, Tower 2 for Oak Ridge National Laboratory, Tower 4 for Advanced Neutron Source Reactor, High Flux Isotope Reactor and Radiochemical Engineering Development Center, Tower 5 for Y- 12, and Tower 6 for the proposed TSS location) at the 10 meter height and are contained in a technical report (Health Risk Data, HNUS 1995a). Table E.2.6-2.-Direction, Distance, and Meteorological Dispersion to Various Maximum Individual Receptors at Oak Ridge Reservation Site Boundary Direction Distance Maximum Receptor For: Atmospheric Dispersion (m) Chi/Q (sec/m3) Release Point: K-25 Incinerator WSW. 2894. K-25 Incinerator 6.4x10-7 E. 9784 Oak Ridge National Laboratory 3039 Stack 7.1x10-8 ENE. 12748 Y-12 Plant 7.8x10-8 E. 9993 High Flux Isotope Reactor Stack 6.9x10-8 SSE. 2307 Tritium Supply Site 5.3x10-7 E. 9916 Advanced Neutron Source Reactor 7.0x10-8 Release Point: X-10 ORNL W. 8756 K-25 Incinerator 5.7x10-8 ENE. 3933 Oak Ridge National Laboratory 3039 Stack 6.1x10-7 NNE. 8943 Y-12 Plant 9.7x10-8 ENE. 4083 High Flux Isotope Reactor Stack 5.7x10-7 WSW. 5407 Tritium Supply Site 2.3x10-7 ENE. 4023 Advanced Neutron Source Reactor 5.9x10-7 Release Point: Y-12 Plant WSW. 15037 K-25 Incinerator 7.2x10-8 S. 5508 Oak Ridge National Laboratory 3039 Stack 4.8x10-8 NNE. 887 Y-12 Plant 1.5x10-6 S. 5676 High Flux Isotope Reactor Stack 4.6x10-8 SW. 12769 Tritium Supply Site 9.6x10-8 S. 5625 Advanced Neutron Source Reactor 4.7x10-8 Release Point: HFIR Stack W. 9896 K-25 Incinerator 4.9x10-8 NE. 3389 Oak Ridge National Laboratory 3039 Stack 7.0x10-7 NNE. 9225 Y-12 Plant 8.0x10-8 ENE. 3440 High Flux Isotope Reactor Stack 7.5x10-7 W. 6431 Tritium Supply Site 9.2x10-8 NE. 3412 Advanced Neutron Source Reactor 7.0x10-7 Release Point: Tritium Supply Site W. 5750 K-25 Incinerator 5.5x10-8 E. 6817 Oak Ridge National Laboratory 3039 Stack 1.5x10-7 NE. 10558 Y-12 Plant 1.9x10-7 E. 7012 High Flux Isotope Reactor Stack 1.5x10-7 SW. 2903 Tritium Supply Site 1.1x10-6 E. 6939 Advanced Neutron Source Reactor 1.5x10-7 Release Point: ANS Reactor W. 10875 K-25 Incinerator 4.2x10-8 NE. 2098 Oak Ridge National Laboratory 3039 Stack 1.5x10-6 NNE. 8104 Y-12 Plant 9.7x10-8 ENE. 2156 High Flux Isotope Reactor Stack 1.6x10-6 W. 7526 Tritium Supply Site 7.3x10-8 ENE. 2124 Advanced Neutron Source Reactor 1.6x10-6 Source: HNUS 1995a. Doses given in this section are associated with 1 year of operation because regulatory standards are given as annual limits. The health effects presented are for a 40-year operational period. E.2.6.1 No Action Atmospheric Releases. For No Action, five of the areas have radioactive releases into the atmosphere from normal operation. Table E.2.6.1-1 presents the estimated annual atmospheric radioactive releases. Table E.2.6.1-1.-Annual Atmospheric Radioactive Releases from Normal Operation of No Action at Oak Ridge Reservation (curies) [Page 1 of 2] Isotope ANSR K-25 ORNL REDC Y-12 Incinerator Plant H-3 7.2x103 0 1.6x104 13 0 Be-7 0 0 6.7x10-6 0 0 Co-57 0 4.1x10-7 0 0 0 Co-60 2.2x10-8 0 2.0x10-4 0 0 Br-82 0 0 8.8x10-4 0 0 Kr-83m 0 0 0 1.9 0 Kr-85 1.8x103 0 4.8x103 230 0 Kr-85m 0 0 0 3.8 0 Kr-87 0 0 0 23 0 Kr-88 0 0 0 17 0 Kr-89 0 0 0 330 0 Kr-90 0 0 0 700 0 Rb-88 0 0 0 13 0 Rb-89 0 0 0 52 0 Rb-90 0 0 0 160 0 Rb-90m 0 0 0 15 0 Sr-89 0 0 0 0.016 0 Sr-90 2.2x10-6 0 3.3x10-5 5.0x10-5 0 Sr-91 0 0 0 0.21 0 Sr-92 0 0 0 1.0x10-5 0 Y-91 0 0 0 7.0x10-4 0 Y-92 0 0 0 1.0x10-5 0 Zr-95 0 0 0 2.0x10-8 0 Nb-95 0 1.5x10-6 0 0 0 Tc-99 0 0.040 0 0 0 Ru-103 0 0 0 6.0x10-8 0 Ru-106 0 0 0 6.0x10-8 0 Ag-110m 0 0 0 1.0x10-9 0 Te-132 1.0x10-5 0 0 0 0 I-129 3.7x10-6 0 4.2x10-6 6.0x10-4 0 I-131 0.13 0 2.1x10-6 0.28 0 I-132 5.4x10-3 0 0 3.8x10-2 0 I-133 0.13 0 0 6.6x10-3 0 I-134 0 0 0 0.23 0 I-135 0.11 0 0 0.030 0 Xe-131m 0 0 0 0.12 0 Xe-133 8.5x103 0 0 380 0 Xe-133m 0 0 0 0.23 0 Xe-135 0 0 0 100 0 Xe-135m 0 0 0 0 0 Xe-137 0 0 0 4.0x103 0 Xe-138 0 0 0 1.7x103 0 Cs-134 0 0 0 2.0x10-9 0 Cs-136 0 0 0 9.0x10-9 0 Cs-137 4.4x10-6 5.5x10-4 6.9x10-5 1.1x10-3 0 Cs-138 0 0 0 400 0 Cs-139 0 0 0 1.5x103 0 Ba-139 0 0 0 92 0 Ba-140 0 0 0 0.22 0 Ba-141 0 0 0 2.0x10-5 0 La-140 3.0x10-6 0 0 0 0 La-141 0 0 0 1.0x10-6 0 Ce-141 0 0 0 5.0x10-8 0 Ce-144 0 0 4 6.0x10-8 0 Nd-147 0 0 0 2.0x10-8 0 Eu-154 0 0 0 4.0x10-10 0 Eu-155 0 0 0 4.0x10-10 0 Eu-156 0 0 0 5.0x10-8 0 Tb-160 0 0 0 8.0x10-10 0 Os-191 0 0 4.0 0 0 Pb-212 0.057 0 0.22 0 0 Th-228 8.8x10-9 2.7x10-3 4.2x10-8 0 0 Th-230 1.1x10-8 8.4x10-5 1.3x10-8 0 0 Th-232 7.8x10-9 1.5x10-5 9.6x10-9 0 0 Th-234 0 0.047 0 0 0 U-234 1.6x10-8 0.011 5.3x10-8 0 0.04 U-235 0 4.6x10-4 0 0 1.2x10-3 U-236 0 4.0x10-11 0 0 1.6x10-4 U-238 0 0.013 0 0 5.5x10-3 Np-237 0 8.1x10-4 0 0 0 Pu-238 0 4.3x10-8 7.3x10-8 4.0x10-9 0 Pu-239 4.3x10-10 5.4x10-5 3.6x10-7 5.0x10-10 0 Pu-240 0 0 0 9.0x10-10 0 Pu-241 0 0 0 1.0x10-7 0 Pu-242 0 0 0 3.0x10-12 0 Cm-244 0 0 0 2.0x10-7 0 Cf-252 0 0 0 1.0x10-8 0 Tables E.2.6.1-2 and E.2.6.1-3 include the atmospheric radiological impacts to the maximally exposed member of the public and the offsite population within 50 miles, respectively. The maximally exposed individual would receive an annual dose of 3.9 mrem. An estimated fatal cancer risk of 7.8x10-5 would result from 40years of operation. The population within 50 miles would receive a dose of 39person-rem in the year 2030 (midlife of operation). An estimated 0.78 fatal cancers could result from 40years of operation. Table E.2.6.1-2.-Doses and Resulting Health Effect to the Maximally Exposed Individual from Atmospheric Releases Associated with Normal Operation at Oak Ridge Reservation Alternative Dose by Pathway (mrem per year) - Percent of Estimated Background 40-Year Fatal Cancer Risk - Inhalation Ingestion Plume Ground Shine Committed - - Immersion Effective Dose Equivalent (mrem per year) No Action - Advanced Neutron 0.34 3 0.55 0.015 3.9 1.3 7.8x10-5 Source Reactor No Action - High Flux Isotope 1.8 8.4x10-3 0.034 5.3x10-4 1.8 0.59 3.6x10-5 Reactor No Action Contribution to Tritium 0.58 0.16 0.11 2.5x10-3 0.85 0.28 1.7x10-5 Supply Site Maximum Receptor from Advanced Neutron Source Reactor No Action Contribution to Tritium 0.57 0.017 0.092 1.3x10-3 0.67 0.22 1.3x10-5 Supply Site Maximum Receptor from High Flux Isotope Reactor Heavy Water Reactor 0.22 3.3 0.013 3.6x10-4 3.5 1.1 7.0x10-5 Modular High Temperature Gas- 0.14 1.9 3.2x10-3 1.5x10-3 2 0.66 4.0x10-5 Cooled Reactor AP600 Reactor 0.13 2.2 0.043 3.8x10-3 2.4 0.78 4.8x10-5 Advanced Boiling Water Reactor 0.27 4.3 0.16 0.017 4.7 1.5 9.4x10-5 CE System 80+ 0.33 4.9 0.015 3.5x10-4 5.2 1.7 1.0x10-4 Simplified Boiling Water Reactor 0.23 3.7 0.039 0.043 4 1.3 8.0x10-5 Full Accelerator Production of 4.8x10-3 0.065 4.2x10-4 4.7x10-6 0.07 0.023 1.4x10-6 Tritium with helium-3 Full Accelerator Production of 0.099 1.3 4.2x10-4 4.7x10-6 1.4 0.46 2.8x10-5 Tritium with spallation-induced lithium conversion Phased Accelerator Production of 4.8x10-3 0.065 1.7x10-4 2.4x10-7 0.070 0.023 1.4x10-6 Tritium Tritium Recycling 0.19 2.6 1.8x10-6 0 2.8 0.92 5.6x10-5 Table E.2.6.1-3.-Doses and Resulting Health Effect to the Population Within 50 Miles of Oak Ridge Reservation from Atmospheric Releases Associated with Normal Operation Alternative Dose by Pathway in 2030 (person-rem) - Percent of Estimated Background 40-Year Fatal Cancers - Inhalation Ingestion Plume Ground Shine Committed - - Immersion Effective Dose Equivalent in 2030 (person-rem) No Action - Advanced Neutron 29 5.4 4.1 0.16 39 0.012 0.77 Source Reactor No Action - High Flux Isotope 27 0.23 3.9 0.055 31 9.5x10-3 0.63 Reactor AP600 Reactor 2.1 7.1 0.52 0.061 9.8 3.0x10-3 0.20 Advanced Boiling Water Reactor 4.3 14 0.84 0.21 19 5.6x10-3 0.37 CE System 80+ 5.3 15 0.16 5.5x10-3 21 6.2x10-3 0.41 Simplified Boiling Water Reactor 3.7 12 0.25 0.66 17 5.0x10-3 0.33 Heavy Water Reactor 3.4 10 0.12 5.6x10-3 14 4.2x10-3 0.28 Modular High Temperature Gas- 2.2 5.7 0.03 0.018 7.9 2.4x10-3 0.16 Cooled Reactor Full Accelerator Production of 1.5 4 4.0x10-3 1.7x10-7 5.5 1.7x10-3 0.11 Tritium with spallation-induced lithium conversion Full Accelerator Production of 0.077 0.2 4.0x10-3 1.7x10-7 0.28 8.6x10-5 5.7x10-3 Tritium with helium-3 Phased Accelerator Production of 0.077 0.2 2.0x10-3 8.6x10-8 0.28 8.6x10-5 5.6x10-3 Tritium Tritium Recycling 3.1 7.9 2.0x10-9 0 11 3.3x10-3 0.22 Liquid Releases. For No Action, some areas may have radioactive releases to the offsite surface water from normal operation. Table E.2.6.1-4 presents the estimated annual liquid radioactive releases. Table E.2.6.1-4.-Annual Liquid Releases from Normal Operation of No Action at Oak Ridge Reservation (curies) Isotope Oak Ridge National Laboratory Y-12 Plant H-3 2.5x103 0 Co-60 0.17 0 Sr-90 3.2 6.4x10-3 Tc-99 0 9.9x10-5 Cs-137 1.8 0 Ra-226 0 0.032 Ra-228 0 1.4 U-234 0 0.11 U-235 0 6.4x10-3 U-238 0 0.12 Np-237 0 9.8x10-4 Th-228 0 6.3x10-3 Th-230 0 2.8x10-3 Th-232 0 4.6x10-4 Th-234 0 0.015 Pu-238 0 0 Pu-239 0 2.3x10-4 Source: ORR 1993a:8. Tables E.2.6.1-5 and E.2.6.1-6, respectively, present the radiological impacts to the maximally exposed individual and the offsite populations using surface water within 50 miles downstream of ORR. The maximally exposed member of the public would receive an annual dose of 14 mrem. An estimated fatal cancer risk of 2.7x10-4 would result from 40years of operation. The population would receive a dose of 18person-remrem in the year 2030. An estimated 0.36 fatal cancers could result from 40years of operation. Table E.2.6.1-5.-Doses and Resulting Health Effect to the Maximally Exposed Individual at Oak Ridge Reservation from Liquid Releases Associated with Normal Operation Alternative Dose by Pathway (mrem per year) Committed Percent of - Effective Dose Background Equivalent (mrem per year) - Fish Ingestion Other Food Drinking Boating Swimming Shoreline Estimated - - Ingestion Water Risk of 40-Year Fatal Cancers No Action - 13 0.022 0.4 4.4x10-5 8.8x10-5 0.027 14 4.4 2.7x10-4 Advanced Neutron Source Reactor No Action - High 13 0.022 0.4 4.4x10-5 8.8x10-5 0.027 14 4.4 2.7x10-4 Flux Isotope Reactor Heavy Water 0.13 4.4x10-3 0.019 1.2x10-5 2.4x10-5 4.8x10-4 0.15 0.05 3.1x10-6 Reactor AP600 0.13 0.06 0.2 6.0x10-6 1.2x10-5 9.4x10-4 0.39 0.13 7.8x10-6 Advanced Boiling 0.041 0.03 0.098 2.8x10-6 5.6x10-6 3.3x10-4 0.17 0.055 3.4x10-6 Water Reactor CE System 80+ 0.1 0.016 0.051 5.3x10-6 1.1x10-5 8.5x10-4 0.18 0.058 3.5x10-6 Simplified Boiling 0.044 0.039 0.13 2.8x10-6 5.6x10-6 3.4x10-4 0.21 0.07 4.3x10-6 Water Reactor Table E.2.6.1-6.-Doses and Resulting Health Effects to the Population Downstream of Liquid Releases Associated with Normal Operation at Oak Ridge Reservation Alternative Dose by Pathway (person-rem in 2030) - Percent of Estimated Background 40-Year Fatal Cancers - Fish Ingestion Drinking Water Boating Swimming Shoreline Committed - - Effective Dose Equivalent in 2030 (person-rem) No Action - 8.8 9.1 2.0x10-3 1.7x10-3 0.22 18 5.8x10-3 0.36 Advanced Neutron Source Reactor No Action - High 8.8 9.1 2.0x10-3 1.7x10-3 0.22 18 5.8x10-3 0.36 Flux Isotope Reactor Heavy Water 0.25 0.43 5.6x10-4 4.8x10-4 3.8x10-3 0.69 2.2x10-4 0.014 Reactor AP600 0.31 4.6 2.8x10-4 2.4x10-4 7.6x10-3 4.9 1.6x10-3 0.099 Advanced Boiling 0.084 2.3 1.3x10-4 1.1x10-4 2.6x10-3 2.4 7.7x10-4 0.048 Water Reactor CE System 80+ 0.2 1.2 2.4x10-4 2.1x10-4 6.8x10-3 1.4 4.5x10-4 0.028 Simplified Boiling 0.089 3 1.3x10-4 1.1x10-4 2.7x10-3 3.1 1.0x10-3 0.062 Water Reactor Worker Doses. Based on measured values during 1991 and 1992 (Dose Reports for 1991 and 1992) and including the contribution from the Advanced Neutron Source Reactor, it is estimated that the average dose to a badged worker involved in No Action activities at ORR in the year 2010 and beyond would equal 17.4 mrem. It is projected that in the year 2010 and beyond, there would be 18,315 badged workers involved in No Action activities. The annual dose among all these workers would equal 319person-rem. From 40years of operation, an estimated fatal cancer risk of 2.8x10-4 would result to the average worker and 5.1 fatal cancers could result among all workers. E.2.6.2 Tritium Supply Technologies and Recycling Atmospheric Releases. For the tritium supply technologies and recycling facilities, total site radiological impacts resulting from tritium supply technologies and recycling facilities can be found by adding the impacts resulting from No Action facilities to those resulting from each tritium supply technology and recycling facility. For example, to determine the radiological impact for the addition of the HWR at ORR, the No Action facilities impacts would be summed with the HWR doses (which includes tritium target extraction) and the tritium recycling dose. Estimated annual atmospheric radioactive releases for the tritium supply technologies and recycling facilities are given in section E.2.3. Tables E.2.6.1-2 and E.2.6.1-3 present the atmospheric radiological impacts by tritium supply technology and recycling facility. The annual dose from total site operation associated with the different tritium supplies ranged from 4.3 to 5.0 mrem to the maximally exposed member of the public and from 50 to 70person-remrem to the 50-mile population in the year 2030. The health effects from 40years of operation are included in both tables. Liquid Releases. There are three technologies that would release liquid discharges to the surface water surrounding ORR. These are the HWR and the Large and Small ALWRs. The liquid releases for these two technologies are given in section E.2.3. For example, to determine the liquid radiological impact for the addition of an HWR at ORR, the No Action liquid impacts must be summed with the HWR liquid doses. Tables E.2.6.1-5 and E.2.6.1-6 present the liquid radiological impacts for the applicable alternatives. The annual doses from total site operations associated with the different tritium supplies that have liquid releases are all approximately 14 mrem to the maximally exposed member of the public, and ranged from 18 to 23 person-rem to the downstream population in the year 2030. The health effects from 40years of operation are included in both tables. E.2.7 Radiological Impacts at Pantex Plant This section presents the radiological impacts of the tritium supply technologies and recycling facilities at Pantex. Section E.2.7.1 presents the radiological releases and resulting impacts from facilities associated with No Action. Section E.2.7.2 presents the radiological releases and resulting impacts from the facilities associated with tritium supply technologies and recycling facilities. For purposes of radiological impact modeling, Pantex was divided into four areas which will release radioactivity in the year 2010. All release points in each area were aggregated into a single release point. Table E.2.7-1 presents the characteristics of each of the release points including location, release height, and minimum distance and annual average dispersion to the site boundary in each of 16directions. In order to calculate the maximum site boundary dose, the dose from each release point to the maximum receptor associated with each of the other release points has been calculated. For example, the dose resulting from releases from Building 12-44 Cell 1 and the proposed TSS has been determined from the maximum receptor from the Burning Ground. Figure E.2.7-1 illustrates the location of each maximum receptor in relation to each release point. The maximum site boundary dose is then determined by the maximum dose of each of the maximum receptors. Table E.2.7-2 presents the distance, direction, and atmospheric dispersion from each release point to each of the maximum receptors. Annual radiological releases were assumed to remain constant during the 40-year operational period. Descriptions of population and food stuff distributions centered on each release area are provided in a technical report Health Risk Data. The joint frequency distribution used for the dose assessment was based on the meteorological measurements for the year 1989 from the National Weather Service at the 10 meter height and is contained in a technical report (Health Risk Data). Doses given in this section are associated with 1 year of operation because regulatory standards are given as annual limits. The health effects presented are for the 40-year operational period. Figure (Page E-56) Figure E.2.7-1.-Location of Maximum Receptors at Pantex Plant. Table E.2.7-1.-Release Point Characteristics, Direction, Distance and Chi/Q at Pantex Plant Boundary Release Tritium Supply Site Burning Ground Building 12-44, Cell 1 Block C Point Latitude 35o19'54.60" 35o20'40.78" 35o18'24.09" 35o18'8.55" Longitude -101o35'15.15" -101o35'4.25" -101o33'25.59" -101o33'30.52" Release Ground Level Ground Level Ground Level Ground Level Height: - - - Accidents - - - - - - Direction Distance Chi/Q 50 Percentile 95 Percentile Distance Chi/Q Distance Chi/Q Distance Chi/Q (m) (sec/m3) Chi/Q Chi/Q (m) (sec/m3) (m) (sec/m3) (m) (sec/m3) (s/m3) (s/m3) N. 2,352 5.7x10-7 8.0x10-7 1.0x10-4 932 2.7x10-6 5,176 1.7x10-7 5,615 1.5x10-7 NNE. 2,397 3.4x10-7 9.3x10-7 1.0x10-4 958 1.6x10-6 2,910 2.5x10-7 5,711 9.3x10-8 NE. 2,822 2.2x10-7 4.8x10-7 7.6x10-5 1,119 9.9x10-7 2,258 3.1x10-7 5,688 7.7x10-8 ENE. 4,177 7.2x10-8 2.3x10-7 4.8x10-5 1,654 3.1x10-7 2,012 2.2x10-7 4,763 5.9x10-8 E. 4,245 8.0x10-8 2.9x10-7 5.3x10-5 3,983 8.8x10-8 1,872 2.8x10-7 4,649 7.0x10-8 ESE. 4,317 5.3x10-8 5.2x10-7 5.7x10-5 4,062 5.8x10-8 1,844 2.0x10-7 4,088 5.8x10-8 SE. 5,047 5.5x10-8 3.9x10-7 4.8x10-5 5,036 5.5x10-8 1,904 2.4x10-7 3,352 1.0x10-7 SSE. 6,165 3.1x10-8 2.4x10-7 3.6x10-5 6,934 2.6x10-8 2,578 1.1x10-7 3,244 7.8x10-8 S. 6,227 5.7x10-8 1.0x10-7 2.6x10-5 7,472 4.4x10-8 2,607 2.1x10-7 3,283 1.5x10-7 SSW. 5,121 4.6x10-8 1.7x10-7 3.9x10-5 5,630 4.0x10-8 2,998 1.0x10-7 3,782 7.3x10-8 SW. 3,348 7.7x10-8 4.0x10-7 7.2x10-5 3,676 6.7x10-8 4,285 5.3x10-8 2,870 9.7x10-8 WSW. 2,801 1.5x10-7 1.8x10-6 1.1x10-4 3,087 1.3x10-7 5,636 5.3x10-8 2,406 1.9x10-7 W. 2,724 1.5x10-7 1.2x10-6 1.1x10-4 2,999 1.3x10-7 5,495 5.4x10-8 2,338 1.9x10-7 WNW. 2,769 1.1x10-7 6.1x10-7 1.0x10-4 1,742 2.4x10-7 5,575 4.1x10-8 2,377 1.5x10-7 NW. 2,860 1.7x10-7 5.6x10-7 9.0x10-5 1,151 7.6x10-7 6,284 5.5x10-8 2,790 1.8x10-7 NNW. 2,411 2.3x10-7 7.2x10-7 9.8x10-5 959 1.1x10-6 5,297 7.1x10-8 4,056 1.0x10-7 Source: HNUS 1995a. Table E.2.7-2.-Direction, Distance, and Meteorological Dispersion to Various Maximum Individual Receptors at the Pantex Site Boundary Direction Distance Maximum Receptor For: Atmospheric Dispersion (m) (m) Chi/Q (s/m3) Release Point: Tritium Supply Site N. 2372 Burning Ground 5.6x10-7 ESE. 4585 Bldg. 12-44 Cell 1 4.9x10-8 N. 2351 Tritium Supply Site 5.7x10-7 NE. 3158 Block A 4.9x10-8 SW. 4272 Block C 5.6x10-7 Release Point: Bldg. 12-44 Cell 1 NNW. 5713 Burning Ground 6.4x10-8 NE. 1834 Bldg. 12-44 Cell 1 4.3x10-7 NNW. 5845 Tritium Supply Site 6.2x10-8 N. 5213 Block A 4.3x10-7 W. 5518 Block C 6.4x10-8 Release Point: HE Burning Ground N. 932 Burning Ground 2.7x10-6 SE. 5066 Bldg. 12-44 Cell 1 5.5x10-8 NNW. 972 Tritium Supply Site 1.1x10-6 ENE. 2039 Block A 5.5x10-8 SW. 5588 Block C 2.7x10-6 Release Point: Block A NNW. 3893 Burning Ground 1.1x10-7 E. 3893 Bldg. 12-44 Cell 1 9.1x10-8 NNW. 4332 Tritium Supply Site 9.5x10-8 N. 3803 Block A 2.7x10-7 WSW. 5183 Block C 5.9x10-8 Release Point: Block C N. 5664 Burning Ground 1.5x10-7 ENE. 4896 Bldg. 12-44 Cell 1 5.7x10-8 N. 5633 Tritium Supply Site 1.5x10-7 NNE. 6162 Block A 8.3x10-8 W. 2339 Block C 1.9x10-7 Source: HNUS 1995a. E.2.7.1 No Action Atmospheric Releases. For No Action, two of the areas have radioactive releases into the atmosphere from normal operation. Table E.2.7.1-1 presents the estimated annual atmospheric radioactive releases for No Action. Liquid Releases. There are no radioactive liquid releases into the offsite environment associated with No Action. Tables E.2.7.1-2 and E.2.7.1-3 include the radiological impacts to the maximally exposed individual and the offsite population within 50 miles, respectively. The maximally exposed individual would receive an annual dose of 1.3x10-3 mrem. An estimated fatal cancer risk of 2.6x10-8 would result from 40years of operation. The population within 50miles would receive a dose of 5.7x10-4 person-rem in the year 2030 (midlife of operation). An estimated 1.1x10-5 fatal cancers could result from 40years of operation. Worker Doses. Based on measured values during the time period from 1989 to 1992 (Twenty-Second Annual Report Radiation Exposure for DOE and DOE Contractor Employees - 1989 DOE/EH-0286P)) and subsequent yearly dose reports), the annual average dose to a badged worker at Pantex was calculated to be 15 mrem. It is projected that in the year 2010 and beyond, there would be 2,465 badged workers involved in No Action activities at Pantex (PX 1993a:2). The annual average dose to these workers was assumed to remain at 15 mrem; the annual total dose among all these workers would then equal 37 person-rem. From 40years of operation, an estimated fatal cancer risk of 2.4x10-4 would result to the average worker and 0.59 fatal cancers could result among all workers. Table E.2.7.1-1.-Estimated Annual Atmospheric Radioactive Releases from Normal Operation of No Action at Pantex Plant (curies) Isotope Weapons Assembly/ Disassembly High Explosive - Building 12-44 Burning Plutonium Cell 1 Ground Storage Helium-3 0.12 0 0 Uranium-238 0 2.1x10-5 0 Source: PX 1993a:2. Table E.2.7.1-2.-Doses and Resulting Health Effect to the Maximally Exposed Individual Resulting from Normal Operation at Pantex Plant Alternative Dose by Pathway (mrem per year) - Percent of Estimated Background 40-Year Fatal Cancer Risk - Inhalation Ingestion Plume Ground Shine Committed - - Immersion Effective Dose Equivalent (mrem per year) No Action 1.3x10-3 5.0x10-6 1.2x10-12 9.9x10-8 1.3x10-3 4.2x10-4 2.6x10-8 No Action contribution to tritium 5.0x10-4 2.9x10-6 4.4x10-13 3.8x10-8 5.0x10-4 1.5x10-4 1.0x10-8 supply site maximum receptor Heavy Water Reactor 0.52 1.9 6.8x10-3 2.4x10-3 2.4 0.7 4.8x10-5 Modular High Temperature 0.07 0.94 1.7x10-3 7.9x10-4 1 0.29 2.0x10-5 Gas-Cooled Reactor Advanced Boiling Water Reactor 0.21 3.2 0.095 9.8x10-3 3.5 1 7.0x10-5 CE System 80+ 0.21 3.1 7.9x10-3 4.2x10-3 3.2 0.93 6.4x10-5 Simplified Boiling Water Reactor 0.21 3.2 0.024 0.024 3.4 0.99 6.8x10-5 AP600 0.21 3.1 0.022 6.4x10-3 3.3 0.96 6.6x10-5 Full Accelerator Production of 2.4x10-3 0.033 1.8x10-4 3.7x10-7 0.035 0.01 7.0x10-7 Tritium with helium-3 Full Accelerator Production of 0.049 0.66 1.8x10-4 3.7x10-7 0.71 0.21 1.4x10-5 Tritium with spallation-induced lithium conversion Phased Accelerator Production of 2.4x10-3 0.033 8.9x10-5 1.9x10-7 0.035 0.01 7.0x10-7 Tritium Tritium Recycling 0.099 1.3 9.4x10-11 0 1.4 0.4 2.8x10-5 Table E.2.7.1-3.-Doses and Resulting Health Effect to the Population Within 50 Miles Resulting from Normal Operation of Pantex Plant Alternative Dose by Pathway (person-rem in 2030) - Percent of Estimated Background 40-Year Fatal Cancers - Inhalation Ingestion Plume Ground Shine Committed - - Immersion Effective Dose Equivalent in 2030 (person-rem) No Action 4.5x10-4 1.2x10-4 5.4x10-13 3.6x10-8 5.7x10-4 5.7x10-4 1.1x10-5 Heavy Water Reactor 0.92 17 8.5x10-3 4.3x10-3 19 0.019 0.37 Modular High Temperature 0.12 6.3 2.1x10-3 1.1x10-3 6.5 6.5x10-3 0.13 Gas-Cooled Reactor Advanced Boiling Water Reactor 0.38 26 0.057 0.016 28 0.028 0.55 CE System 80+ 0.38 25 0.01 7.5x10-3 26 0.026 0.51 Simplified Boiling Water Reactor 0.39 24 0.017 0.041 25 0.025 0.49 AP600 0.38 25 0.033 0.012 26 0.026 0.51 Full Accelerator Production of 4.3x10-3 0.22 2.5x10-4 8.3x10-9 0.23 2.3x10-4 4.6x10-3 Tritium with helium-3 Full Accelerator Production of 0.087 4.4 2.5x10-4 8.3x10-9 4.5 4.5x10-3 0.09 Tritium with spallation-induced lithium conversion Phased Accelerator Production of 4.3x10-3 0.22 1.3x10-4 4.2x10-9 0.23 2.3x10-4 4.6x10-3 Tritium Tritium Recycling 0.17 8.8 1.7x10-10 0 9 9.0x10-3 0.18 E.2.7.2 Tritium Supply Technologies and Recycling Total site radiological impacts resulting from tritium supply and recycling can be found by adding the impacts resulting from No Action facilities to those resulting from tritium supply technologies and recycling facilities. For example, to determine the radiological impact for the addition of the HWR at Pantex, the No Action facilities impacts would be summed with the HWR doses (which includes tritium target extrication) and the tritium recycling doses. Estimated annual atmospheric radioactive releases for the tritium supply technologies and recycling facilities are given in section E.2.3. Tables E.2.7.1-2 and E.2.7.1-3 present the radiological impacts by tritium supply technology and recycling facility. There are no radioactive liquid releases into the offsite environment associated with any tritium supply and recycling alternative. The annual doses from total site operations associated with the different tritium supplies ranged from 1.4 to 4.9 mrem to the maximally exposed individual and from 9.2 to 37person-rem to the 50-mile population in the year 2030. The health effects from 40years of operation are included in both tables. E.2.8 Radiological Impacts at Savannah River Site This section presents the radiological impacts of the tritium supply technologies and recycling facilities at SRS. Section E.2.8.1 presents the radiological releases and resulting impacts from facilities associated with No Action. Section E.2.8.2 presents the radiological releases and resulting impacts from the facilities associated with the tritium supply technologies and recycling facilities. For purposes of radiological impact modeling, SRS was divided into seven areas. All potential release points in each area were aggregated into a single release point. Table E.2.8-1 presents the characteristics of each of the release points including location, release height, minimum distance, and annual average dispersion to the site boundary in each of 16directions. In order to calculate the maximum site boundary dose, the dose from each release point to the maximum receptor associated with each of the other release points has been calculated. For example, the dose resulting from releases from F- , H- , and S-Areas, the K- and L-Reactors, and the proposed TSS is determined for the maximum receptor from the Savannah River Tech Center Laboratory in A-Area. Figure E.2.8-1 illustrates the location of each maximum receptor in relation to each release point. The maximum site boundary dose is then determined by the maximum dose of each of the maximum receptors. Table E.2.8-2 presents the direction, distance, and atmospheric dispersion from each release point to each of the maximum receptors. Annual radiological releases were assumed to remain constant during the 40-year operational period. Descriptions of population and food stuffs distributions centered on each release area are provided in the technical report Health Risk Data. The joint frequency distribution used for the dose assessment was based on the meteorological measurements for the year 1985 from the meteorological tower at SRS at the 61-meter height and is contained in the technical report Health Risk Data. Figure (Page E-63) Figure E.2.8-1.-Location of Maximum Receptors at Savannah River Site. Table E.2.8-1.-Release Point Characteristics, Direction, Distance, and Chi/Q at Savannah River Site Boundary [Page 1 of 2] - A-Area F-Area H-Area K-Area Latitude 33o20'24 36" 33o17'11.40" 33o17'11.04" 33o12'42.12" Longitude -81o44'6.72" -81o40'34.68" -81o38'25.07" -81o39'49.31" Release 10.0 meters 10.0 meters 10.0 meters 10.0 meters Height: - - - - - - - Accident - - Direction Distance Chi/Q Distance Chi/Q Distance Chi/Q 50 Percentile 95 Percentile Distance Chi/Q (m) (sec/m3) (m) (sec/m3) (m) (sec/m3) Chi/Q Chi/Q (m) (sec/m3) (sec/m3) (sec/m3) N. 1,920 2.6x10-6 10,895 2.3x10-7 12,280 1.9x10-7 5.6x10-6 6.3x10-5 19,112 1.1x10-7 NNE. 3,260 1.5x10-6 12,646 2.2x10-7 12,860 2.2x10-7 4.9x10-6 9.0x10-5 21,418 1.1x10-7 NE. 5,447 8.2x10-7 14,762 2.1x10-7 14,889 2.1x10-7 3.9x10-6 5.8x10-5 21,727 1.2x10-7 ENE. 12,422 3.0x10-7 18,536 1.7x10-7 15,959 2.1x10-7 3.4x10-6 6.5x10-5 15,630 2.2x10-7 E. 21,490 1.6x10-7 17,109 2.2x10-7 14,055 2.8x10-7 5.2x10-6 8.0x10-5 15,628 2.5x10-7 ESE. 23,855 9.9x10-8 16,944 1.6x10-7 13,690 2.1x10-7 5.0x10-6 8.8x10-5 13,427 2.1x10-7 SE. 27,208 4.8x10-8 19,773 7.3x10-8 17,634 8.6x10-8 3.9x10-6 8.1x10-5 11,434 1.5x10-7 SSE. 25,922 3.3x10-8 18,932 5.0x10-8 17,662 5.5x10-8 2.1x10-6 5.0x10-5 10,834 1.1x10-7 S. 14,864 7.1x10-8 18,520 5.3x10-8 18,108 5.5x10-8 8.9x10-7 6.9x10-5 11,121 1.1x10-7 SSW. 7,327 2.7x10-7 15,478 9.6x10-8 18,479 7.6x10-8 2.3x10-6 5.7x10-5 10,683 1.6x10-7 SW. 5,280 6.9x10-7 11,521 2.3x10-7 14,338 1.7x10-7 3.5x10-6 5.4x10-5 10,613 2.6x10-7 WSW. 3,434 1.5x10-6 9,639 3.5x10-7 14,207 2.1x10-7 3.6x10-6 6.0x10-5 9,147 3.8x10-7 W. 2,588 1.8x10-6 9,421 3.0x10-7 12,768 2.0x10-7 5.5x10-6 6.8x10-5 8,856 3.3x10-7 WNW. 1,751 2.5x10-6 9,838 2.2x10-7 12,662 1.5x10-7 4.8x10-6 9.4x10-5 12,320 1.6x10-7 NW. 1,575 2.3x10-6 9,459 1.8x10-7 11,893 1.3x10-7 4.9x10-6 1.0x10-4 13,277 1.1x10-7 NNW. 1,395 2.7x10-6 9,963 1.7x10-7 11,750 1.4x10-7 5.3x10-6 5.8x10-5 17,101 8.3x10-8 - L-Area S-Area Tritium Supply Site Latitude 33o12'38.52" 33o17'42.72" 33o15'28.80" Longitude -81o37'26.40" -81o38'35.15" -81o37'33.96" Release 10.0 meters 10.0 meters 10.0 meters Height: - - - - - - - Accident Direction Distance Chi/Q Distance Chi/Q Distance Chi/Q 50 Percentile 95 Percentile (m) (sec/m3) (m) (sec/m3) (m) (sec/m3) Chi/Q Chi/Q (sec/m3) (sec/m3) N. 20,731 9.6x10-8 11,286 2.2x10-7 15,564 1.4x10-7 4.0x10-6 4.7x10-5 NNE. 21,141 1.1x10-7 11,984 2.4x10-7 15,856 1.6x10-7 3.7x10-6 6.9x10-5 NE. 15,520 1.9x10-7 14,240 2.2x10-7 16,560 1.8x10-7 3.4x10-6 5.0x10-5 ENE. 12,052 3.1x10-7 15,663 2.2x10-7 13,279 2.7x10-7 4.4x10-6 8.1x10-5 E. 13,327 3.1x10-7 14,623 2.7x10-7 11,822 3.6x10-7 6.7x10-6 9.9x10-5 ESE. 11,184 2.7x10-7 14,223 2.0x10-7 11,822 3.6x10-7 4.7x10-6 8.3x10-5 SE. 9,886 1.9x10-7 18,443 8.1x10-8 14,551 1.1x10-7 5.1x10-6 1.0x10-4 SSE. 9,298 1.3x10-7 18,667 8.1x10-8 14,255 7.4x10-8 2.8x10-6 6.5x10-5 S. 9,587 1.3x10-7 19,113 5.1x10-8 14,819 7.2x10-8 1.2x10-6 8.9x10-5 SSW. 12,157 1.3x10-7 19,048 7.3x10-8 16,910 8.5x10-8 2.6x10-6 6.3x10-5 SW. 12,497 2.1x10-7 14,544 1.7x10-7 15,675 1.5x10-7 3.1x10-6 4.8x10-5 WSW. 13,516 2.2x10-7 12,868 2.4x10-7 13,864 2.1x10-7 3.7x10-6 6.2x10-5 W. 12,508 2.0x10-7 12,473 2.0x10-7 14,146 1.7x10-7 4.8x10-6 6.0x10-5 WNW. 15,662 1.1x10-7 11,487 1.7x10-7 14,384 1.3x10-7 4.1x10-6 8.0x10-5 NW. 17,091 8.2x10-8 10.982 1.5x10-7 14,974 9.8x10-8 3.6x10-6 7.9x10-5 NNW. 19,321 7.0x10-8 10,741 1.5x10-7 15,132 9.7x10-8 3.8x10-6 4.3x10-5 Source: HNUS 1995a. Doses given in this section are associated with 1year of operation because regulatory standards are given as annual limits. The health effects presented are for the 40-year operational period. Table E.2.8-2.-Direction, Distance, and Meteorological Dispersion to Various Maximum Individual Receptors at Savannah River Site Boundary [Page 1 of 2] Direction Distance Maximum Receptor For: Atmospheric Dispersion (m) Chi/Q (sec/m3) Release Point: A- Area NNW. 1,396 A-Area 2.7x10-6 SSW. 9,638 F-Area 2.1x10-7 ESE 24,209 H-Area 9.7x10-8 S. 16,255 K-Area 6.3x10-8 ESE. 24,650 L-Area 9.5x10-8 E. 23,833 S-Area 1.4x10-7 ESE. 24,175 Tritium Supply Site 9.7x10-8 Release Point: F-Area NW. 9,460 A-Area 1.8x10-7 WSW. 9,638 F-Area 3.5x10-7 E. 17,358 H-Area 2.1x10-7 SW. 12,787 K-Area 2.0x10-7 ESE. 17,299 L-Area 1.5x10-7 E. 17,704 S-Area 2.1x10-7 ESE. 16,980 Tritium Supply Site 1.5x10-7 Release Point: H-Area NW. 11,932 A-Area 1.3x10-7 W. 12,930 F-Area 1.9x10-7 E. 14,055 H-Area 2.8x10-7 SW. 15,055 K-Area 1.6x10-7 ESE. 14,139 L-Area 2.0x10-7 E. 14,357 S-Area 2.8x10-7 ESE. 13,760 Tritium Supply Site 2.1x10-7 Release Point: K Reactor NNW. 17,101 A-Area 8.3x10-8 WNW. 12,320 F-Area 1.6x10-7 ENE. 16,962 H-Area 1.9x10-7 WSW. 9,146 K-Area 3.8x10-7 E. 15,627 L-Area 2.5x10-7 ENE. 18,642 S-Area 1.7x10-7 ENE. 15,803 Tritium Supply Site 2.1x10-7 Release Point: L Reactor NW. 19,081 A-Area 7.1x10-8 WNW. 15,661 F-Area 1.1x10-7 ENE. 13,579 H-Area 2.6x10-7 W. 12,775 K-Area 2.0x10-7 ENE. 12,051 L-Area 3.1x10-7 NE. 15,519 S-Area 1.9x10-7 ENE. 12,302 Tritium Supply Site 3.0x10-7 Release Point: S-Area NW. 11,159 A-Area 1.5x10-7 WSW. 12,867 F-Area 2.4x10-7 ESE. 14,517 H-Area 1.9x10-7 SW. 15,548 K-Area 1.5x10-7 ESE. 14,759 L-Area 1.9x10-7 E. 14,622 S-Area 2.7x10-7 ESE. 14,328 Tritium Supply Site 1.9x10-7 Release Point: Tritium Supply Site NW. 14,974 A-Area 9.8x10-8 W. 14,146 F-Area 1.7x10-7 E. 12,498 H-Area 2.1x10-7 WSW. 14,285 K-Area 2.1x10-7 E. 12,001 L-Area 3.5x10-7 ENE. 13,487 S-Area 2.6x10-7 E. 11,822 Tritium Supply Site 3.6x10-7 Source: HNUS 1995a. E.2.8.1 No Action Atmospheric Releases. For No Action, all of the areas have radioactive releases into the atmosphere from normal operation. Tables E.2.8.1-1 and E.2.8.1-2 present the estimated annual atmospheric radioactive releases. Tables E.2.8.1-3 and E.2.8.1-4 present the radiological impacts to the maximally exposed individual and the offsite population within 50 miles, respectively. The maximally exposed individual would receive an annual dose of 2.8mrem. An estimated fatal cancer risk of 5.6x10-5 would result from 40years of operation. The population within 50miles would receive a dose of 250person-rem in the year 2030 (midlife of operation). An estimated 4.9fatal cancers could result from 40years of operation. Table E.2.8.1-1.-Annual Atmospheric Radioactive Releases from Normal Operation of No Action at Savannah River Site (curies) Isotope SRTC K-Reactor L-Reactor F-Canyon H-Canyon RBOF H-3 0 3.9x103 670 0 0 0 C-14 0 0 0 0 0 0 Ar-41 0 25 0 0 0 0 Cr-51 0 0 0 0 0 0 Co-60 0 0 0 0 0 3.6x10-7 Se-79 0 0 0 0 0 0 Sr-89 0 0 0 0 0 0 Sr-90 0 0 0 2.6x10-5 1.1x10-4 0 Y-90 0 0 0 0 0 0 Y-91 0 0 0 0 0 0 Zr-95 0 0 0 0 0 0 Nb-95 0 0 0 0 0 0 Tc-99 0 0 0 0 0 0 Ru-106 0 0 0 0 0 0 Rh-106 0 0 0 0 0 0 Sn-126 0 0 0 0 0 0 Sb-125 0 0 0 0 0 0 Te-125m 0 0 0 0 0 0 Te-127m 0 0 0 0 0 0 Te-127 0 0 0 0 0 0 I-129 0 0 0 0 0 0 I-131 5.7x10-5 1.4x10-7 0 0 2.1x10-5 0 I-133 1.2x10-3 0 0 0 0 0 I-135 0.13 0 0 0 0 0 Xe-135 2.4x10-3 0 0 0 0 0 Cs-134 0 0 0 0 0 0 Cs-135 0 0 0 0 0 0 Cs-137 3.9x10-6 4.1x10-7 2.6x10-6 8.5x10-6 8.7x10-5 1.7x10-6 Ce-144 0 0 0 0 0 0 Pr-144 0 0 0 0 0 0 Pm-147 0 0 0 0 0 0 Sm-151 0 0 0 0 0 0 Eu-152 0 0 0 0 0 0 Eu-154 0 0 0 0 0 0 Eu-155 0 0 0 0 0 0 U-235 0 0 0 7.3x10-4 3.7x10-5 0 Pu-238 0 0 0 4.1x10-5 1.8x10-4 0 Pu-239 0 0 0 1.6x10-4 5.6x10-5 0 Pu-240 0 0 0 0 0 0 Pu-241 0 0 0 0 0 0 Am-241 2.3x10-8 0 0 2.4x10-4 2.4x10-5 0 Cm-244 7.9x10-8 0 0 3.6x10-6 7.5x10-6 0 Note: SRTC - Savannah River Technology Center; RBOF - Receiving Basin for Offsite Fuels. Source: SRS 1993a:3. Table E.2.8.1-2.-Annual Atmospheric Radioactive Releases from Waste Management Facilities During Normal Operation of No Action at Savannah River Site (curies) Isotope Waste Management - - - - F-Area H-Area Tritium DWPF Consolidated Facilities Incineration Facility H-3 0 2.1x10-5 3.5x104 20 1.2x103 C-14 0 0 0 0.021 0 Ar-41 0 0 0 0 0 Cr-51 0 0 0 0 0.015 Co-60 0 0 0 6.1x10-8 1.4x10-4 Se-79 0 0 0 8.8x10-9 0 Sr-89 0 0 0 0 6.0x10-4 Sr-90 0 0 0 2.3x10-5 0.022 Y-90 0 0 0 2.4x10-5 7.6x10-5 Y-91 0 0 0 0 4.5x10-4 Zr-95 0 0 0 0 4.7x10-4 Nb-95 0 0 0 0 1.5x10-3 Tc-99 0 0 0 3.8x10-7 0 Ru-106 0 1.8x10-6 0 3.2x10-5 1.8x10-4 Rh-106 0 0 0 0 1.8x10-4 Sn-126 0 0 0 6.9x10-8 0 Sb-125 0 0 0 6.7x10-7 0 Te-125m 0 0 0 1.0x10-5 0 Te-127m 0 0 0 4.5x10-9 0 Te-127 0 0 0 4.4x10-9 0 I-129 0 0 0 8.2x10-5 0 I-131 0 0 0 0 0 I-133 0 0 0 0 0 I-135 0 0 0 0 0 Xe-135 0 0 0 0 0 Cs-134 0 3.7x10-8 0 2.9x10-5 0 Cs-135 0 0 0 9.4x10-7 0 Cs-137 5.0x10-6 2.0x10-5 0 4.1x10-3 2.4x10-4 Ce-144 0 1.2x10-7 0 3.0x10-6 2.3x10-4 Pr-144 0 0 0 3.1x10-6 2.3x10-4 Pm-147 0 0 0 7.6x10-6 9.1x10-4 Sm-151 0 0 0 1.6x10-7 0 Eu-152 0 0 0 1.4x10-9 0 Eu-154 0 0 0 2.3x10-7 0 Eu-155 0 0 0 1.6x10-7 0 U-235 0 0 0 0 0 Pu-238 0 0 0 7.9x10-7 1.4x10-4 Pu-239 0 0 0 7.1x10-9 5.2x10-7 Pu-240 0 0 0 4.8x10-9 0 Pu-241 0 0 0 7.7x10-7 0 Am-241 0 0 0 8.6x10-9 0 Cm-244 0 0 0 2.7x10-8 0 Note: DWPF - Defense Waste Processing Waste Facility. Source: SRS 1993a:3. Table E.2.8.1-3.-Doses and Resulting Health Effect to Maximally Exposed Individual from Atmospheric Releases Associated with Normal Operation at Savannah River Site Alternative Dose by Pathway (mrem per year) - Percent of Estimated Background 40-Year Fatal Cancer Risk - Inhalation Ingestion Plume Ground Shine Committed - - Immersion Effective Dose Equivalent (mrem per year) No Action 0.2 2.6 2.8x10-4 6.9x10-5 2.8 0.89 5.6x10-5 No Action with Tritium Upgrade 0.15 1.8 2.8x10-4 6.9x10-5 2 0.63 4.0x10-5 Heavy Water Reactor 0.059 0.93 2.2x10-3 1.0x10-4 0.99 0.31 2.0x10-5 Modular High Temperature 0.036 0.5 5.3x10-4 3.3x10-4 0.53 0.17 1.1x10-5 Gas-Cooled Reactor Advanced Boiling Water Reactor 0.076 1.2 0.012 3.8x10-3 1.3 0.42 2.6x10-5 CE System 80+ 0.094 1.4 2.8x10-3 1.0x10-4 1.5 0.46 2.9x10-5 Simplified Boiling Water Reactor 0.065 1.1 4.0x10-3 0.012 1.1 0.36 2.2x10-5 AP600 0.034 0.6 9.4x10-3 1.1x10-3 0.64 0.2 1.3x10-5 Full Accelerator Production of Tritium 1.2x10-3 0.017 6.2x10-5 6.7x10-12 0.018 5.7x10-3 3.6x10-7 with helium-3 Full Accelerator Production of Tritium 0.024 0.33 6.2x10-5 6.7x10-12 0.35 0.11 7.0x10-6 with spallation-induced lithium conversion Phased Accelerator Production 1.2x10-3 0.017 3.2x10-5 3.4x10-12 0.018 5.7x10-3 3.6x10-7 of Tritium Tritium Recycling 0.14 2 1.4x10-10 0 2.1 0.67 4.2x10-5 Tritium Recycling Upgrade 0.088 1.2 7.9x10-11 0 1.3 0.41 2.6x10-5 Table E.2.8.1-4.-Doses and Resulting Health Effect to the Population Within 50 Miles of Savannah River Site from Atmospheric Releases Associated with Normal Operation Alternative Dose by Pathway (person-rem in 2030) - Percent Estimated of 40-Year Background Fatal Cancers - Inhalation Ingestion Plume Immersion Ground Shine Committed - - Effective Dose Equivalent in 2030 (person-rem) No Action 13 230 6.8x10-3 8.0x10-3 250 0.11 4.9 No Action with Recycle Upgrade 10 160 6.8x10-3 8.0x10-3 180 0.075 3.5 Heavy Water Reactor 3.3 84 0.051 4.6x10-3 87 0.037 1.7 Modular High Temperature 2.1 40 0.01 0.013 42 0.018 0.84 Gas-Cooled Reactor Advanced Boiling Water Reactor 4.2 110 0.36 0.16 110 0.048 2.2 CE System 80+ 5 120 0.089 4.8x10-3 120 0.053 2.5 Simplified Boiling Water Reactor 3.5 85 0.13 0.58 89 0.038 1.8 AP600 2 60 0.32 0.056 62 0.027 1.2 Full Accelerator Production of Tritium 0.079 1.4 8.9x10-4 4.9x10-11 1.5 6.4x10-4 0.03 with helium-3 Full Accelerator Production of Tritium with 1.5 28 8.9x10-4 4.9x10-11 29 0.012 0.58 spallation-induced lithium conversion Phased Accelerator Production of Tritium 0.079 1.4 4.5x10-4 2.5x10-11 1.5 6.4x10-4 0.03 Tritium Recycling 9 170 8.7x10-9 0 180 0.077 3.6 Tritium Recycling Upgrade 5.6 100 5.3x10-9 0 110 0.047 2.2 Liquid Releases. For No Action, some areas may have radioactive releases to the offsite surface water from normal operation. Table E.2.8.1-5 presents the estimated annual liquid radioactive releases. Table E.2.8.1-5.-Annual Liquid Radioactive Releases from Normal Operation of No Action at Savannah River Site (curies) Isotope Savannah River Site Release Isotope Savannah River Site Release H-3 1.4x103 Cs-137 0.079 S-35 3.2x10-5 Pm-147 2.7x10-4 Co-60 4.8x10-5 U-235 1.4x10-5 Sr-90 9.0x10-3 Pu-238 1.3x10-5 Y-91 2.0x10-5 Pu-239 3.5x10-4 Sb-125 7.2x10-5 Am-241 6.9x10-7 Cs-134 2.2x10-5 Cm-244 2.0x10-7 Source: SRS 1993a:3. Tables E.2.8.1-6 and E.2.8.1-7 present the radiological impacts to the maximally exposed member of the public and the offsite populations using water from the Savannah River downstream of SRS to the Atlantic Ocean. The maximally exposed member of the public would receive an annual dose of 0.077mrem. An estimated fatal cancer risk of 1.5x10-6 would result from 40 years of operation. The population would receive a dose of 0.45person-rem in the year 2030. An estimated 9.0x10-3fatal cancers could result from 40 years of operation. Table E.2.8.1-6.-Doses and Resulting Health Effect to the Maximally Exposed Member of the Public from Liquid Releases Associated with Normal Operation at Savannah River Site Alternative Dose by Pathway (mrem per year) Committed Percent of - Effective Dose Background Equivalent (mrem per year) - Fish Other Food Drinking Boating Swimming Shoreline - - Estimated Ingestion Ingestion Water 40-Year Fatal Cancers No Action 0.065 2.6x10-3 8.4x10-3 6.5x10-7 1.3x10-6 1.9x10-4 0.077 0.024 1.5x10-6 Heavy Water 0.076 1.9x10-3 8.4x10-3 2.3x10-6 4.6x10-6 2.0x10-4 0.086 0.027 1.7x10-6 Reactor Advanced Boiling 0.023 0.017 0.057 7.0x10-7 1.4x10-6 1.5x10-4 0.097 0.031 1.9x10-6 Water Reactor CE System 80+ 0.057 6.6x10-3 0.023 2.1x10-6 4.3x10-6 3.7x10-4 0.087 0.027 1.7x10-6 Simplified Boiling 0.023 0.017 0.057 7.0x10-7 1.4x10-6 1.5x10-4 0.097 0.031 1.9x10-6 Water Reactor AP600 0.075 0.026 0.086 2.1x10-6 4.2x10-6 4.1x10-4 0.19 0.060 3.8x10-6 Table E.2.8.1-7.-Doses and Resulting Health Effect to the Population from Liquid Releases Associated with Normal Operation Downstream of Savannah River Site Facility Dose by Pathway (person-rem in 2030) - Percent of - Background - Fish Drinking Boating Swimming Shoreline Committed - Estimated Ingestion Water Effective Dose 40-Year Equivalent Fatal Cancers in 2030 (person-rem) No Action 0.16 0.29 7.1x10-6 2.1x10-6 3.6x10-4 0.45 2.0x10-4 9.0x10-3 Heavy Water Reactor 0.28 0.22 2.6x10-5 7.4x10-6 3.9x10-4 0.5 2.2x10-4 0.01 Advanced Boiling Water 0.12 2.0 7.6x10-6 2.2x10-6 2.8x10-4 2.1 9.5x10-4 0.042 Reactor CE System 80+ 0.25 0.75 2.4x10-5 6.8x10-6 7.1x10-4 1.0 4.5x10-4 0.02 Simplified Boiling 0.12 2.0 7.6x10-6 2.2x10-6 2.8x10-4 2.1 9.5x10-4 0.042 Water Reactor AP600 0.29 3.0 2.3x10-5 6.6x10-6 7.7x10-4 3.3 1.5x10-3 0.066 Worker Doses. It is projected that the annual average dose among the 14,864 badged workers involved in No Action activities at SRS in the year 2010 and beyond would be 32 mrem and the annual total dose among all these workers would equal 480person-rem (SRS1993a:3). From 40 years of operation, an estimated fatal cancer risk of 5.2x10-4would result to the average worker and 7.7fatal cancers could result from all workers. E.2.8.2 Tritium Supply Technologies and Recycling Atmospheric Releases. For the tritium supply technologies and recycling facilities, total site radiological impacts resulting from tritium supply technologies and recycling facilities can be found by adding the impacts resulting from No Action facilities to those resulting from tritium supply technologies and recycling facilities. The No Action doses include contributions from current tritium extraction and recycle operations. For the population doses, these contributions are 30person-rem and 180person-rem, respectively. For the maximally exposed individual, they are 0.35mrem and 1.2mrem, respectively. As an example, to determine the radiological impact for the addition of an HWR at SRS, the No Action facilities doses would be summed with the HWR doses (which includes tritium target extraction) and the tritium recycling doses. Estimated annual atmospheric radio- active releases for the tritium supply and recycling facilities are given in section E.2.3. Tables E.2.8.1-3 and E.2.8.1-4 present the radiological impacts by tritium supply technology and recycling facility. The annual dose from total site operations associated with the different tritium supplies ranged from 2.5 to 3.9mrem to the maximally exposed individual and from 220 to 340 person-rem to the 50-mile population in the year 2030. The health effects from 40 years of operation are included in both tables. Liquid Releases. The HWR and the Large and Small ALWRs technologies would release liquid discharges to the surface water surrounding SRS. The liquid releases for these technologies are presented in section E.2.3. For example, to determine the liquid radiological impact for the addition of the HWR at SRS, the No Action liquid radiological impacts would be summed with the HWR liquid doses. Tables E.2.8.1-5 and E.2.8.1-6 present the liquid radiological impacts for the applicable alternatives. The annual incremental doses from total site operation associated with the different tritium supplies that would release liquids ranged from 0.086to 0.19 mrem to the maximally exposed member of the public and from 0.5 to 3.3person-rem to the downstream population in the year 2030. The health effects from 40 years of operation are included in both tables. E.3 Hazardous Chemical Impacts to Human Health E.3.1 Background There are two types of adverse health effects normally assessed for hazardous chemical exposure. These are carcinogenic and noncarcinogenic effects. For this reason, two kinds of tables were developed (i.e., a table for chemical toxicity profiles (E.3.2-1) and a table of exposure limits (E.3.3-1)). Table E.3.2-1 characterizes each chemical (either found at the site or associated with tritium supply and recycling) qualitatively so that the risk assessor can determine how exposure might occur (e.g., route of exposure), what effect might occur (e.g., central nervous system dysfunction), and whether the chemical might possess other properties that might affect its bioavailability in a given matrix (e.g., air, water, or soil). Table E.3.3-1 provides the risk assessor with the necessary information to calculate the risk or expected adverse effects that would be expected when an individual is exposed to a hazardous chemical for a long term at low levels (chronic exposure) or to higher concentrations for a short term (acute) exposure. Where a dose effect calculation is required (mg/kg/day), the Reference Dose is applicable and where an inhaled concentration effect is required, the Reference Concentration (mg/m3) is applicable for chronic exposures. The Permissible Exposure Limit, Threshold Limit Value, Short Term Exposure Limit, and Reference Exposure Limit values regulate worker exposures over short-term (e.g., 15-min to 8-hour) exposures that are allowed for these time intervals without adverse acute effects. It was assumed that under normal operation conditions, members of the public would only receive chronic exposures at low levels from the air emissions from a centrally located source term at each site; because the chemicals are not released into water or soil, that exposure is entirely from inhalation. It was further assumed that the maximally exposed member of the public would be at the site boundary and this assumption was used for all public exposure calculations. For worker exposures to hazardous chemicals, it was assumed that individuals were exposed only to low air emission concentrations during an 8-hour day for a 40-hour week at a point 100meters from a centrally located source term, since precise placement of source terms onsite cannot be made. Further, it cannot be determined where various workers would be relative to the emission sources. For every site involved in the analyses, Hazard Indexes (HI) were calculated for all alternative actions relevant to the site. The concentrations of hazardous chemicals identified with each option have been developed by modeling each chemical concentration and individually comparing these concentrations to the Reference Concentration (unique to each chemical) to yield Hazard Quotients (HQ) and the HQs summed to yield the HIs for public exposure. For worker exposures, the concentrations are compared to the Permissible Exposure Limit (unique to each chemical) to yield HQs that are summed to yield HIs. The slope factors for all carcinogens are multiplied by the inhaled dose to determine the cancer risk. Table E.3.3-1 has all slope factors, Reference Concentrations, Reference Dose values, and Permissible Exposure Limits that were used in deriving HIs and cancer risks. E.3.2 Chemical Toxicity Profiles Table E.3.2-1 provides the reader with pertinent facts about each chemical that will be included in the risk assessment. This includes the Chemical Abstracts Service number, which aids in a search for information available on any specific chemical and ensures a positive identity regardless of which name or synonym is used. It also contains physical information (i.e., solubility, vapor pressure, and flammability) as well as presenting incompatibility data that is useful in determining whether a hazard might exist and the nature of the hazard. The route of exposure, target organs, and carcinogenicity provide an abbreviated summary on how individuals may get exposed, what body functions could be affected, and whether chronic exposure could lead to cancer in an exposed population. Table E.3.2-1 presents information described above for the hazardous chemicals analyzed in this PEIS. Table E.3.2-1.-Chemical Toxicity Profiles [Page 1 of 13] Compound Chemical Solubility* Vapor Flammabilitya,b Incompatibilitiesa Route of Target Carcinogenicityd Abstracts Pressure** Exposurea,c Organsa Service No. Acetic acid 64-19-7 Misciblea 11 mma Class II Strong oxidizers (esp. inh Respiratory system, f chromic acid, sodium skin, eyes, teeth peroxide and nitric acid), and strong caustics Acetone 67-64-1 Misciblea 180 mma Class IB Oxidizers, acids inh, ing, Respiratory system, f con skin Acetonitrile 75-05-8 Misciblea 73 mma Class IB Strong oxidizers inh, abs, Kidneys, liver, e ing, cardiovascular con system, CNS, lungs, skin, eyes Acetylene 74-86-2 Quitek,l 40 atml Flammablel Brass and copper, salts, inhi Lungs (not serious) f soluble in oxidants, and silver water (I vol./I vol.) Acrolein 107-02-8 40%a 210 mma Class IB Oxidizers, acids, inh, ing, Heart, eyes, skin, EPA Group C: Possible alkalis, ammonia, con respiratory system human carcinogeng amines Acrylonitrile 107-13-1 7%a 83 mma Class IB Strong oxidizers, acids, inh, ing, CVS, liver, kidneys, EPA Group B1: and alkalis, bromine abs, CNS, skin, brain Probable human amines con tumor, lung, and carcinogen; limited bowel cancer evidence in human studiesg Aliphatic alcohol 71-36-3 9%a 6mma Class ICa Strong oxidizersa inh, Eyes, skina, respiratory f (n-butyl alcohol) absa, system ing, con Aliphatic hydrocarbons 107-13-1 Insoluble 98mma Class IAa Oxidizersa inh, ing, Eyes, respiratory f (e.g., cyclohexane) (77 F) con system, skin, CNS Ammonia 7664-41-7 34% in >1 atma Combustible Strong oxidizers, acids, inh, ing, Respiratory system, f water at gas, but halogens, salts of con eyes 68 Fa difficult to silver and zinc burn Antimony 7440-36-0 Insoluble Approx.a Noncombustible Strong oxidizers inh, con Respiratory system, EPA Group A based on 0 mm solid in bulk acidsa, halogenated CVS, skin, eyes radionuclide form, but a acids carcinogenicity moderate explosion hazard in the form of dust when exposed to flame Aromatic petroleum 50-32-8 Insolublel q q e inh, ingl Respiratory systeml, EPA Group B2: distillate in water skin Probable human (benzo[a]pyrene) carcinogen Aromatic hydrocarbons benzo (a) pyrene 50-32-8 Insoluble e e e ing, inhl Skin, respiratory, liver, EPA Group B2 intestine, colonl chrysene 218-01-9 Insoluble e e e ing, inhl intestine, colon EPA Group B2 pyrene 129-00-0 Insoluble e e e inh, ingj Kidneyl EPA Group D Benzene 71-43-2 0.07%a 75 mma Class IB Strong oxidizers, many inh, abs, Blood, CNS, skin, EPA Group A: Human fluorides and ing, bone marrow, eyes, carcinogeng perchlo-rates, nitric con respiratory system acid Bromoform 75-25-2 0.1%a 5mma Noncombustible Lithium, sodium inh, abs, Skin, liver, kidneys, EPA Group B2: potassium, calcium, ing respiratory system, Probable human alumi-num, zinc, CNS carcinogen; sufficient magnesium, strong evidence from animal caustics, acetone studies, inadequate evidence or no data from human studiesg 2-Butoxyethanol 111-76-2 Misciblea 0.8mma Class IIIAa Strong Oxidizers; and inh, Liver, kidneya, f strong causticsa absa, lymphoid system ing, skin, blood, eyes, con respiratory system n-Butyl acetate 123-86-4 1.0%a 15 mma Class IC Nitrates; strong inh, ing, Eyes, skin, respiratory f (77 F) oxidizers, alkalis, con system acids n-Butyl alcohol 71-36-3 9.0%a 6 mma Class IC Strong oxidizersa inh, abs, Skin, eyes, respiratory f ing, systema con Cadmium 7440-43-9 Insolublea Approx Noncombustible Strong oxidizers; ele- inh, ing Respiratory system, EPA Group B1: 0 mma solid in bulk mental sulfur, kidneys, prostate, Probably human form, but will selenium, and blood carcinogen; limited burn in powder tellurium evidence in human form studiesg Cadmium oxide 1306-19-0 Insolublea Approx Noncombustible e inh Respiratory system, f 0 mma kidneys, blood Carbon tetrachloride 56-23-5 0.05%a 91 mma Noncombustible Chemically-active inh, abs, CNS, eyes, lungs, liver, EPA Group B2: metals such as ing, kidneys, skin Sufficient evidence in sodium, potas-sium con animals, no data for and magnesium; humans fluorine; aluminum; Note: forms highly toxic phosgene gas when exposed to flames or welding arcs Chlorine 7782-50-5 0.7%a >1 atma Noncombustible Reacts explosively or inh, con Respiratory system f gas, but a forms explosive com- strong oxidizer pounds with many common substances such as acetylene, ether, ammonia Chlorobenzene 108-90-7 0.05%a 12 mm Class IC Strong oxidizers inh, ing, Respiratory system, EPA Group D: Not (77 con eyes, skin, CNS, liver classified as to F)a human carcinogenicityg Chlorodifluoromethane 75-45-6 e e Nonflammablel, e inhm Respiratoryj system, EPA Group C:i (Freon 22) combustible CNS, liver, kidney, Possible human spleen carcinogen Chloroform 67-66-3 0.5% 160 mma Noncombustible Strong caustics, inh, ing, Liver, kidneys, heart, EPA Group B2: (77 F)a chemically active con eyes, skin Probable human metals (aluminum or carcinogen; sufficient magne-sium powder, evidence from animal sodium and studies, inadequate potassium), strong evidence or no data oxidizers from human studiesg Cleaning solvent 138-22-7 Slightly 0.4 tors Autoignition at e inh Respiratory system f (butyl lactate) soluble in (20 C) 382.2 C water Cyanogen 460-19-5 NA 1.8mml Flammablel e inh, conl Respiratoryl system, f Lower eyes explosive limit = 6.6% Upper limit = 32% Cyclohexane 110-82-7 Insolublea 98 mm Class IA Oxidizers inh, ing, Eyes, respiratory f (77 con system, skin, CNS F)a Diacetone alcohol 123-42-2 Misciblea 1mma Class IIa e inh, Liver, respiratorya f absa, system, CNS ing depression, eyes Dibutyl phthalate 84-74-2 0.5% <0.01 Class IIIB Nitrates; strong inh, ing, Respiratory system, GI EPA Group D: Not (77 F)a mma oxidizers, alkalis and con tract classified as to acids; liquid chlorine human carcinogenicityg 1,3-Dichloropropene 542-75-6 e 28 tors 28.3 to 3.5 C e inh, con Eyes, lung, kidney, EPA Group B2: (20 C) Flashpoint liver Probable human explosive carcinogen; sufficient limits: 5 to evidence from animal 145% in air studies, inadequate evidence or no data from human studiesg Diethylene glycol 111-90-0 NA 4.62mml Flash Pt.l 201 F e ing, Skin, eyesl f ethyl ether absl, con Dimethylformamide 68-12-2 Misciblea 4 mma Class IIIA Carbon tetrachloride, inh, abs, Liver, kidneys, CNS, f (77 F) other halogenated ing, skin compounds when in con contact with iron; strong oxidizers; alkyl aluminum; inorganic nitrates Dioctyl phthalate 117-81-7 Insolublea >0.01 Class IIIB Nitrates; strong inh, con, Eyes, upper respiratory EPA Group B2: mma oxidizers, acids and ing system, GI tract Sufficient evidence in alkalis animals, no data for humans 2,4-Dinitrotoluene 121-14-2 Insolublea 1.0 mm Combustible Strong oxidizers, inh, abs, Blood, liver, EPA Group B2: solid, but caustics, metals such ing, CVS (For mixture)g difficult to as tin and zinc con ignite Epoxy solvent 108-88-3 0.05%a at 20mma at Class IBa Strong oxidizers inh, CNS, livera, kidneys, f (e.g., toluene) 61 F 65 F absa, skin ing, con Ethyl acetate 141-78-6 10% 74 mma Class IB Nitrates; strong inh, ing, Eye, skin, respiratory f (77 F)a oxidizers, alkalis and con system acids Ethyl benzene 100-41-4 0.01%a 10 mm Class IB Strong oxidizers inh, ing, Eyes, upper respiratory EPA Group D: Not (79 con system, skin, CNS classifiable as to F)a human carcinogenicityg Ethanol 64-17-5 Miscible 43 tors Class IA - inh, ing Liver, kidney Group 1., IARCr with (20 C) water and most organic solvents Ethylene oxide 75-21-8 Misciblea >1 atma Class IA Strong acids, alkalis inh, ing, Eyes, blood, IARC Group 2A: and oxidizers; con respiratory system, Suspect human chlorides of iron, liver, CNS, kidneys carcinogens, aluminum and tin; inadequate evidence oxides of iron and in humans, adequate aluminum evidence in animals (EPA Group B2)l,r Formaldehyde 50-00-1 Misciblea >1 atm/ Class IIIB Strong oxidizers, inh, ing, Respiratory system, EPA Group B1: 1 mma alkalis and acids; con eyes, skin Sufficient evidence in phenols; urea; pure animals, limited formaldehyde has a evidence in humansg tendency to polymerize Formic acid 64-18-6 Misciblea 35 mma Class IIIA Strong oxidizers, inh, ing, Respiratory system, - strong caustics, con skin, kidneys, liver, concentrated sulfuric eyes acid Freon 22 75-45-6 e e Nonflammablel e inhm Respiratory, CNS, Human Evidence (chloro- liver, kidney, spleenj Inadequate, Animal difluoromethane) evidence limitedi Freon 12 75-71-8 0.03% >1 atma Nonflammable Chemically active inh, con Cardiovascular f (dichloro- (77 F)a gas metals (i.e., sodium, system, peripheral difluoromethane) potassium, calcium, nervous system pow-dered aluminum, zinc and magnesium)a Freon 114 76-14-2 0.01% >1 atma Nonflammable Chemically active inh, ing, Respiratory system, f (dichloro- gas metals (i.e., sodium, con cardiovascular tetrafluoroethane) potassium, calcium, system pow-dered aluminum, zinc and magnesium; acids; acid fumes Freon 11 75-69-4 0.1010 690 mma e Sodium, potassium, inh, ing, CVS, skin f (trichloro- (77 F)a aluminum, calcium, con fluoromethane) lithium, barium Freon 113 76-13-1 0.02%a 285 mma Flammable Aluminum, barium, inh, CNS, skin f (trichloro- (77 F) lithium, samarium, ing, trifluoroethane) sodium, potassium conl alloy, titanium Freon TF 75-69-4 0.1%a at 690mma N/Aa Aluminum, barium, inh, CNS, skina f (trichloro- 77 F lithium, samarium, inga, trifluoromethane) sodium, potassium con alloy, titanium Heptane 142-82-5 0.005% 40 mm Class IB Strong oxidizers inh, ing, Skin, respiratory Class D (60 F)a (72 con system, peripheral F)a nervous system Hexane 110-54-3 0.002%a 150 mm Class IB Strong oxidizers inh, ing, Skin, eyes, respiratory EPA Group D: Not (77 con system, peripheral classifiable as to F)a nervous systema, s human carcinogenicityg Hydrocarbons 142-82-5 .005%a at 40mma at Class IBa Strong oxidizers inh, Skin, respiratorya EPA Group D: Not (n-heptane) 60 F 72 F inga, system, peripheral classifiable as to con nervous system human carcinogenicity Hydrochloric acid 7647-01-0 67% >1 atma Nonflammable Metals, hydroxides, inh, ing, Respiratory system, f (86 F)a gas amines, alkalis; con skin, eyes corrosive to most metals Hydrofluoric acid 7664-39-3 Misciblea >1 atma Nonflammable Metals, water or steam; inh, abs, Eyes, respiratory f gas corrosive to metals ing, system, skin con Hydrogen cyanide 74-90-8 Misciblea 630mma Class IAa Many (amines, inh, CNSa, cardiovascular, f oxidizers, acids, absa, liver, kidneys water, caustics, etc.) ing, con Hydrogen sulfide 7783-06-4 0.4%a >1 atma Flammable gas Strong oxidizers, inh, ing, Respiratory system, f strong nitric acid, con eyes metals Isobutane 75-28-5 NA 2.01mmp Lowerp e inha Respiratoryp system f (2-methyl propane) flammability limit = 1.8% Upper limit = 48% Isobutyl acetate 110-19-0 0.6%a at 13mma Class IBa Nitrates, strong inh Skin, eyesa, respiratory f 77 F oxidizers, alkalis, and system acids Isopropyl alcohol 67-63-0 Miscible 33 mma Class IB Strong oxidizers, inh, ing, Eyes, respiratory f acetaldehyde, con system, skin chlorine, ethylene oxide, acids, isocyanates Manganese 7439-96-5 Insolublea Approx Combustible Oxidizers inh, ing Respiratory system, EPA Group D: Not 0 mma solid CNS, blood, kidneys classifiable as to human carcinogenicityg Mercury (vapor) 7439-97-6 Insolublea 0.0012 Noncombustible Acetylene, ammonia, inh, Skin, respiratory f mm liquid chlorine dioxide, abs, system, CNS, azides, calcium, con kidneys, eyes sodium carbide, lithium, rubidium, copper Methane 74-828 NA NA Flammable gask NA inh Lungs f Methanol 67-56-1 Misciblea 92 mma Class IB Strong oxidizers inh, abs, Eyes, skin, CNS, GI f ing, tract con Methyl bromide 74-83-9 2%a >1 atma Flammable gas Aluminum, inh, abs, CNS, respiratory f but only in the magnesium, strong ing, system, skin, eyes pres-ence of a oxidizers con high energy ignition source Methylene chloride 75-09-2 2%a 350 mma Combustible Strong oxidizers; inh, ing, Skin, CVS, eyes, CNS EPA Group B2: liquid caustics; chemically- con Sufficient evidence in active metals such as animals, limited aluminum, evidence in humansg magnesium, powders, etc. Methyl ethyl ketone 78-93-3 28%a 71 mma Class IB Strong oxidizers, inh, ing, CNS, lungs EPA Group D: Not (2-butanone) amines, ammonia, con classifiable as to inorganic acids, human caustics, copper, carcinogenicityn isocyanates, pyridines Methyl isobutyl 108-10-1 2%a 16 mma Class IB Strong oxidizers, inh, ing, Respiratory system, e ketone potassium con eyes, skin, CNS tertbutoxide Methyl-tert-butyl-ether 1634-04-4 4%k 245mmk Ignition at 224 Unstable in acid inhk e EPA Group D: Not (2-methoxy-2-methyl- Ck solutionsk classifiable as to propane) human carcinogenicityg Naphtha 8030-30-6 Insolublea <5 atma Class II Strong oxidizers inh, ing, Respiratory system, e con eyes, skin Naphthalene 91-20-3 0.003%a 0.08 mma Combustible Strong oxidizers, inh, abs, Eyes, blood, liver, EPA Group D: Not solid, but will chronic anhydride ing, kidneys, skin, RBC, classifiable as to take effort to con CNS human ignite carcinogenicityg Nickel (refinery dust) 7440-02-0 Insolublea Approx Noncombustible Strong acids, sulfur, inh, ing, Lungs, paranasal sinus, Nickel refinery dust- 0 mma solid in bulk selenium, wood and con CNS EPA Group A: form other combustibles, Human carcinogeng nickel nitrate Nitric acid 7697-37-2 Misciblea 48 mma Noncombustible Combustible materials; inh, ing, Eyes, respiratory f liquid, but metallic powders; con system, skin, teeth increases hydrogen sulfide, flammability carbides; alcohols of combustible materials Nitrocellulose 9004-70-4 NA NA Highly NA NA NA f (cellulose tetranitrite) flammablek Perfluoroalkylether 26675-46- NA NA NA e inhl Respiratory system, f (isoflurane) 7 CNS Perfluoro compounds 76-14-2 0.01%a >1 Atma Nonflammablea Chemically active inh, Respiratorya system, f (e.g., tetra CFCs) metals inga, cardiovascular con Phenol 108-95-2 9% 0.4 mma Combustible Strong oxidizers, inh, abs, Liver, kidneys, skin EPA Group D: Not (77 F)a solid calcium ing, classifiable as to hypochlorite, con human aluminum chloride, carcinogenicityg acids Phosphoric acid 7664-38-2 Misciblea 0.03 mma Noncombustible Strong caustics, most inh, ing, Respiratory system, f liquid metals con eyes, skin Propane 74-98-6 0.01%a >1 Atma Lower limit 2%p e inh, CNSa f Upper limit cona 52% Propyl glycol methyl 107-98-2 NA NA NA e ing, Skinl f ether (propylene glycol conl, monomethyl ether) inn Propylene dichloride 78-87-5 0.3%a 40 mma Class IB Strong oxidizers, inh, con, Skin, eyes, respiratory EPA Group C: strong acids ing system, liver, Possible human kidneys carcinogen Resins and formers 584-84-9 Insolublea 0.01mma Class IIIBa Many (strong inh, Respiratorya system, f (toluene-2,4- at 77 F oxidizers, water, inga, skin diisocyanate) acids, bases, etc.) con Sodium hydroxide 1310-73-2 111%a Approx Noncombustible Water, acids, inh, ing, Eyes, respiratory f 0 mma solid, but flammable liquids, con system, skin when in organic halo-gens, contact with metals (aluminum, water may tin, and zinc), generate nitromethane enough heat to ignite combustible materials Sulfuric acid 7664-93-9 Misciblea 1 mm Noncombustible Organic materials, inh, ing, Respiratory system, f (295 liquid, but can chlorates, carbides, con eyes, skin, teeth F)a ignite finely fulminates, water, divided powdered metals; combustible corrosive to metals metals Tetrachloroethane 79-34-5 0.3%a 9 mm Noncombustible Chemically-active inh, abs, liver, kidneys, CNS EPA Group C: Possible (1,1,2,2) (86 liquidk metalsa, strong ing, human carcinogeng F)a caustics, fuming of con sulfuric acid Tetrachloroethylene 127-18-4 0.02% 9mm Noncombustible Strong oxidizers, ing, Liver, kidneys, CNS, EPA Group B2 @77F 86F liquid chemically active inh, eyes, upper metals, caustic soda, con respiratory system sodium hydroxide Tetrahydrofuran 109-99-9 Misciblea 132mma Class IBa e inh, Eyes, skin, CNSa, f inga, respiratory system con Toluene 108-88-3 0.05%a 20 mma Class IB Strong oxidizers inh, abs, CNS, liver, kidneys, f (61F) (65 F) ing, skin con Trichloroethane 71-55-6 Insolublek e Nonflammable e inhk CNS, eyes, mucus f (1,1,1) liquidk membranek Trichloroethylene 79-01-6 0.1% 58 mm Class IC Strong caustics and inh, ing, Respiratory system, EPA Group B: (77 F)a alkalis; chemically- con heart, liver, kidneys, Sufficient evidence in active metals such as CNS, skin animals, inadequate barium, lithium, evidence in humansn sodium, magnesium, titanium, and beryllium Uranium 7440-61-1 Insolublea Approx Combustible Carbon dioxide, carbon inh, con, Skin, bone marrow, EPA Group A: Human 0 mma solid, esp. tetra-chloride, nitric ing lymphatics Carcinogenn turnings and acid, fluorine powder Vinyl acetate 108-05-4 1g/50ml 100n Flammable Oxygenn, hydrogen inh, Skin, eyem EPA Group D: waterk at liquidm peroxide conm Inadequate evidence 20C in animalso Vinyl chloride 75-01-4 0.1% (77 >1 atma Flammable gas Copper, oxidizers, inh Liver, CNS, blood, EPA Group A: Human F)a aluminum, respiratory system, Carcinogeno peroxides, iron, steel lymphatic system Xylene 1330-20-7 Insolublea 7/9/9 IB (o-) IC (m-, Strong oxidizers inh, abs, CNS, eyes, GI tract, f (o-, m-, mma p-) ing, blood, liver, kidneys, p-isomers) con skina a NIOSH 1990a; ORNL 1993b. b Flammable liquids are classified by OSHA (29 CFR 1910.106) as follows: Class IA-Fl.P below 73 F and BP below 100 F; Class IB-Fl.P below 73 F and BP at or above 100 F; Class IC-Fl.P at or above 73 F and below 100 F; Class II-Fl.P at or above 100 F and below 140 F; Class IIIA-Fl.P at or above 140 F and below 200 F; Class IIIB-Fl.P at or above 200 F. c Routes of exposure abbreviated as follows: inh-inhalation; abs-skin absorption; ing-ingestion; con-skin and/or eye contact. d DHHS 1992a. e Information is not available. f Not applicable, not found to be carcinogenic. g ORNL 1993b. Note: CNS - central nervous system; CVS - cardiovascular system Note: * - % by weight/volume, i.e. g/100ml. Note: ** - at 68 F unless indicated otherwise. h Mixture of compounds, therefore, a representative compound was selected for chemical group. i DHHS 1992b. j ACGIH 1991. k Merck 1989a. l Lewis 1992a. m DHHS 1986a. n EPA 1993c. o CFR 40, EPA 1992. p Patty's Industrial Hygiene and Toxicology. q Moderate explosion hazard in form of dust when exposed to flame. r IARC, International Agency for Research on Cancer. s Encyclopedia of Occupational Health, 1983. E.3.3 Regulated Exposure Limits Hazardous chemicals are regulated by various agencies to provide protection to the public (EPA) and workers (Occupational Safety and Health Administration (OSHA)), while others and National Institute for Occupational Safety and Health (American Conference of Governmental Industrial Hygienists) provide guidelines. The Reference Dose and Reference Concentration set by EPA represent exposure limits for long-term (chronic) exposure at low doses and concentrations, respectively, that can be considered safe from adverse noncancer effects. The Permissible Exposure Limit represents levels set by OSHA that are considered safe for 8-hour exposures so as not to cause noncancer adverse effects. The slope factor for each toxic chemical can be used to convert the daily intake averaged over a lifetime of exposure to the incremental risk of an individual developing cancer. Table E.3.3-1, presents the information on exposure limits that was used to develop HQs for individual hazardous chemicals, the HIs for each option at each site, and the slope factors used to calculate cancer risk for each chemical. Table E.3.3-1.-Regulated Exposure Limits [Page 1 of 9] Compound Chemical Referencea Referencea Cancer Slope Factora Occupational Exposure Levels Abstracts Dose Concentration Classa (mg/kg/day)-1 Service No. (oral) (inhalation) (mg/kg/day) (mg/m3) Acetic acid 64-19-7 0.175 0.613c NA None OSHA-PEL: 25 mg/m3d ACGIH-TLV: 25 mg/m3, 37 mg/m3 [STEL]e NIOSH-REL: same as abovef Acetone 67-64-1 1x10-1 10.5r D None OSHA-PEL: 1,800 mg/m3 and STEL of 2,400 mg/m3d ACGIH-TLV: same as abovee NIOSH-REL: 590 mg/m3 and a IDLH of 48,400 mg/m3d Acetonitrile 75-05-8 6.0x10-3 k 0.021c NA None OSHA-PEL: 70 mg/m3, 105 mg/m3 [15-min. STEL]d [0.050*] ACGIH-TLV: 67 mg/m3, 101 mg/m3 [skin] [STEL]e NIOSH-REL: 34 mg/m3d Acetylene 74-86-2 18.63c 65.22c NA None OSHA-PEL: 2,662 mg/m3f ACGIH-TLV: Simple Asphyxiant NIOSH-REL: 2,708mg/m3f Acrolein 107-02-8 2.0x10-2* 7.0x10-2c C None OSHA-PEL: 0.25 mg/m3 and STEL of 0.8 mg/m3d ACGIH-TLV: 0.23 mg/m3c NIOSH-REL: 0.25 mg/m3d Acrylonitrile 107-13-1 1.0x10-3* 2.0x10-3 k B1k 0.24 OSHA-PEL: 4.42 mg/m3 (inhal) ACGIH-TLV: 4.3 mg/m3 NIOSH: IDLH = 1,105 mg/m3 Aliphatic alcohol 71-36-3 0.1 k 0.35 c D None OSHA-PEL: 150 mg/m3 (skin) (n-butyl alcohol) ACGIH-TLV: 152 mg/m3 (ceiling) NIOSH-REL: 150 mg/m3 (skin) (ceiling) Aliphatic hydrocarbons 110-82-7 7.35 [PNL] 25.7c NA None OSHA-PEL: 1050 mg/m3e (e.g., cyclohexane) ACGIH-TLV: 1030 mg/m3e NIOSH-REL: 1050 mg/m3d Ammonia 7664-41-7 0.0286j 0.1* NA None OSHA-PEL: 27 mg/m3d ACGIH-TLV: 17 mg/m3 and a 15 minute STEL of 27 mg/m3d NIOSH-REL: 18 mg/m3 and a 15 minute STEL of 27 mg/m3d Antimony 7440-36-0 4.0x10-4# 1.4x10-3 c NA None OSHA-PEL: 0.5 mg/m3d ACGIH-TLV: 0.5 mg/m3e NIOSH-REL: 0.5 mg/m3 and IDLH of 80 mg/m3d Aromatic hydrocarbons - - - - - OSHA-PEL: 0.2 mg/m3 benzo (a) pyrene 50-32-8 NA NA B2k 5.79k ACGIH-TLV: suspected human carcinogen Chrysene 218-01-9 NA NA B2k NIOSH-REL: controlled as a carcinogen Pyrene 129-00-0 3x10-2 1.05x10-1 c Dk Aromatic petroleum 50-32-8 1.4x10-3 l 4.4x10-3 c B2 5.79k OSHA-PEL: 0.2 mg/m3 distillate (benzo[a]pyrene) ACGIH-TLV: Suspect Human Carcinogen NIOSH-REL: Controlled as Carcinogen Benzene 71-43-2 0.224b 0.783c A .029 OSHA-PEL: 3.25 mg/m3 and a STEL of 16.25 (oral) ACGIH-TLV: 32 mg/m3 NIOSH-REL: 0.325 mg/m3 and a IDLH of 9750 mg/m3 Bromoform 75-25-2 2x10-2 6.99x10-4 B2k 0.0079k OSHA-PEL: 5 mg/m3d ACGIH-TLV: 5.2 mg/m3e NIOSH-REL: 5 mg/m3d 2-Butoxyethanol 111-76-2 8.47x10-1 b 2.96c NA None OSHA-PEL: 240 mg/m3 ACGIH-TLV: 121 mg/m3 NIOSH-REL: 120 mg/m3 n-Butyl acetate 123-86-4 4.97b 17.4c NA None OSHA-PEL: 710 mg/m3, 950 mg/m3 [15-min. STEL]g ACGIH-TLV: same as aboveh NIOSH-REL: same as abovef n-Butyl alcohol 71-36-3 1.05b 3.675c D None OSHA-PEL: 150 mg/m3 [skin] [ceiling]d ACGIH-TLV: 152 mg/m3 [ceiling]e NIOSH-REL: 150 mg/m3 [skin] [ceiling]d Cadmium 7440-43-9 5x10-4 1.75x10-3 c B1 i OSHA-PEL: 0.2 mg/m3d ACGIH-TLV: 0.05 mg/m3e NIOSH-IDLH: 50 mg/m3d Cadmium oxide 1306-19-0 7x10-3 b 2.5x10-2 c NA None OSHA-PEL: 0.1 mg/m3d ACGIH-TLV: 0.05 mg/m3e NIOSH-REL: 2 mg/m3 and an IDLH of 9 mg/m3d Carbon tetrachloride 56-23-5 7x10-4 2.5x10-3 B2 1.3x10-1 OSHA-PEL: 12.6 mg/m3 [8-hr TWA]d (oral)* ACGIH: 31 mg/m3 [skin]e 5.3x10-2 NIOSH-REL: 12.6 mg/m3 [60-min ceiling limit]d (inhal)* Chlorine 7782-50-5 0.011 3.9x10-2 c NA None OSHA-PEL: 1.5 mg/m3 and a 15-minute STEL of 3 mg/m3d [PNL] ACGIH-TLV: 1.5 mg/m3e NIOSH-REL: 1.5 mg/m3d Chlorobenzene 108-90-7 2x10-2 k 2x10-2* D None OSHA-PEL: 350 mg/m3d (alternate ACGIH-TLV: 46 mg/m3e method) NIOSH-IDLH: 11,232 mg/m3d Chlorodifluoromethane 75-45-6 24.8b 86.7c NA None OSHA-PEL: 3,500 mg/m3 (Freon 22) ACGIH-TLV: 3,540 mg/m3 NIOSH-REL: NA Chloroform 67-66-3 1x10-2 0.03497 B2 0.0061 OSHA-PEL: 240 mg/m3g (oral) OSHA-TWA: 9.78 mg/m3d 0.08 ACGIH-TLV: 49 mg/m3e (inhal) NIOSH-IDLH: 4,960 mg/m3d Cleaning solvent (butyl lactate) 138-22-7 0.208b 0.73c NA NA OSHA-PEL: 30 mg/m3g ACGIH-TLV: 30 mg/m3e Cyanogen 460-19-5 4x10-2 k 0.14c NA None OSHA-PEL: NA ACGIH-TLV: 21 mg/m3 OSHA-TWA: 20 mg/m3 Cyclohexane 110-82-7 7.35b 25.7c NA None OSHA-PEL: 1,050 mg/m3f ACGIH-TLV: 1,030 mg/m3f NIOSH-REL: 1,050 mg/m3f Diacetone alcohol 123-42-2 1.68l 5.88c NA None OSHA-PEL: 240 mg/m3 ACGIH-TLV: 240 mg/m3 NIOSH-REL: 240 mg/m3 Dibutyl phthalate 84-74-2 0.1 0.3497 D None OSHA-PEL: 5 mg/m3d ACGIH-TLV: 5 mg/m3e Dichloropropene (1,3) 542-75-6 3x10-4 2x10-2 B2 0.18 OSHA-PEL: 5 mg/m3g ACGIH-TLV: 4.5 mg/m3e Diesel exhaust i i i i i OSHA-PEL: i ACGIH-TLV:i NIOSH-REL:i Diethylene glycol ethyl ether 111-90-0 2.0* 7.0c NA None OSHA-PEL: NA ACGIH-TLV: NA RTECS: 2,500 mg/kg (rabbit), intravenous LD50; 6600 mg/kg oral LD50 (mouse) Dimethylformamide 68-12-2 0.1* 0.35c NA None NIOSH-REL: 30 mg/m3 [8-hr TWA]d OSHA-PEL: same as aboveg ACGIH-TLV: same as abovee Dinitrotoluene (2,4) 121-14-2 2x10-3 6.9x10-3 B2 None OSHA-PEL: 1.5 mg/m3 (mixture) ACGIH-TLV: 0.15 mg/m3 NIOSH-REL: Reduce to lowest level Dioctyl phthalate 117-81-7 0.02 0.07c B2 0.014 OSHA-PEL: 5 mg/m3, 10 mg/m3 [15-min. STEL]d (Di(2-ethylhexyl) phthalate) (oral) ACGIH-TLV: 5 mg/m3e NIOSH-REL: 5 mg/m3 (skin)d Epoxy solvent 108-88-3 2x10-1 k 4.0x10k NA None OSHA-PEL: 375 mg/m3 (e.g., toluene) ACGIH-TLV: 188 mg/m3 NIOSH-REL: 375 mg/m3 Ethyl benzene 100-41-4 0.1 1 D None OSHA-PEL: 435 mg/m3d ACGIH-TLV: 434 mg/m3e NIOSH-REL: 435 mg/m3d and an IDLH of 8,820 mg/m3d Ethyl acetate 141-78-6 0.9 3.15c NA None OSHA-PEL: 1,400 mg/m3d [9.0*] ACGIH-TLV: same as abovee NIOSH-REL: same as aboved Ethyl alcohol (ethanol) 64-17-5 13.2b 46.1c NA None OSHA-PEL: 1,900 mg/m3g ACGIH-TLV: 1,880 mg/m3e Ethylene oxide 75-21-8 0.0128 0.0447 IARC 0.35 OSHA-PEL: 1.83 mg/m3f EPA B1 (inhal) ACGIH-TLV: 1.8 mg/m3e NIOSH-REL: 0.18 mg/m3 and an IDLH of 1,464 mg/m3f Fiske 604 i i i i i OSHA-PEL: i ACGIH-TLV: i NIOSH-REL: i Formaldehyde 50-00-0 0.2 none B1 4.5x10-2 OSHA-PEL: 1 ppm with a 15 min. STEL of 2 ppmd (inhal)* ACGIH-TLV: 1 ppme unit risk NIOSH-REL: 0.016 ppm with a 15 minute STEL of 0.1 ppm factor: (Ceiling)d 1.3x10-5 (inhal) Formic acid 64-18-6 2 6.994 NA NA OSHA-PEL: 9 mg/m3d ACGIH-TLV: 9.4 mg/m3e NIOSH-REL: 9 mg/m3 and an IDLH of 57.3 mg/m3d Freon 22 75-45-6 24.8b 86.7c NA None OSHA-PEL: 3,500 mg/m3d (chlorodifluoromethane) ACGIH-TLV: 3,540 mg/m3e NIOSH-REL: NA Freon 12 75-71-8 0.2b 0.7* NA None OSHA-PEL: 4,950 mg/m3d (dichlorodifluoromethane) ACGIH-TLV: same as abovee NIOSH-REL: same as aboved Freon 11 75-69-4 0.3 0.7c NA None OSHA-PEL: 5,620 mg/m3, 5620 mg/m3 [ceiling]d (trichlorofluoromethane) [0.2*] ACGIH-TLV: same as abovee NIOSH-REL: same as aboved Freon 113 76-13-1 30 105c NA None OSHA-PEL: 7,600 mg/m3, 9500 mg/m3 [ST]d (trichlorotrifluoroethane) ACGIH-TLV: 7,670 mg/m3, 9590 mg/m3 [STEL]e NIOSH-REL: 7,600 mg/m3, 9500 mg/m3 [STEL]f Freon TF 75-69-4 3.0x10-1 k 1.05c NA None OSHA-PEL: 5,600 mg/m3 (trichlorotrifluoromethane) ACGIH-TLV: 5,620 mg/m3 NIOSH-REL: 5,600 mg/m3 Heptane 142-82-5 11.5b 40.3c D None OSHA-PEL: 1,600 mg/m3, 2000 mg/m3 [STEL]d ACGIH-TLV: 1,640 mg/m3, 2050 mg/m3 [STEL]e NIOSH-REL: 350 mg/m3, 1800 mg/m3 [15-min. ceiling]f Hexane 110-54-3 6x10-2 j 0.2 D None OSHA-PEL: 1,800 mg/m3g ACGIH-TLV: 176 mg/m3e NIOSH-REL: 180 mg/m3 and an IDLH of 17,900 mg/m3d Hydrocarbons (n-heptane) 142-82-5 11.5b 40.25c D None OSHA-PEL: 1,600 mg/m3d ACGIH-TLV: 1,640 mg/m3 NIOSH-REL: 350 mg/m3 Hydrochloric acid 7647-01-0 2x10-3 j 7x10-3 a NA None OSHA-PEL: 7 mg/m3h [PNL] ACGIH-TLV: 7.5 mg/m3 [ceiling]e NIOSH-REL: 7 mg/m3f Hydrofluoric acid 7664-39-3 0.018b 0.063c NA None OSHA-PEL: 2.5 mg/m3, 5 mg/m3 [STEL]d ACGIH-TLV: 2.6 mg/m3 [ceiling]i NIOSH-REL: 2.5 mg/m3, 5 mg/m3 [STEL]g Hydrogen cyanide 74-90-8 2.0x10-2 k 7.0x10-2 c NA None OSHA-PEL: 11 mg/m3 ACGIH-TLV: 11 mg/m3 NIOSH-REL: 5 mg/m3 Hydrogen fluoride 7664-39-3 6x10-2 0.21c NA None OSHA-PEL: 2.5 mg/m3, 5 mg/m3, 15 STEL [PNL] ACGIH-TLV: 2.6 mg/m3e NIOSH-REL: 2.5 mg/m3 and an IDLH of 24.9 mg/m3 Hydrogen sulfide 7783-06-4 3x10-3 k 9x10-4 k NA None OSHA-PEL: 28.4 mg/m3, the TWA is 14 mg/m3d ACGIH-TLV: 14 mg/m3, and the STEL is: 21 mg/m3e NIOSH-REL: 15 mg/m3, the IDLH is 426 mg/m3d Isobutane (2 methyl propane) 75-28-5 10.01m 35.035c NA None OSHA-PEL: NA ACGIH-TLV: NA RTECS-LD50: 1,898 mg/m3s Isobutyl acetate 110-19-0 4.9p 17.15c NA None OSHA-PEL: 300 mg/m3 ACGIH-TLV: 700 mg/m3 NIOSH-REL: 700 mg/m3 Isopropyl alcohol 67-63-0 6.9 [PNL] 24.15c NA None OSHA-PEL: 980 mg/m3 [8-hr TWA],d 1,225 mg/m3, [15-min ST limit]d ACGIH-TLV: same as abovee NIOSH-REL: same as aboved Manganese 7439-96-5 0.14 5x10-3k Dk None OSHA-PEL: 5 mg/m3, TWA = 1 mg/m3d, 3 mg/m3[STEL] ACGIH-TLV: 2 mg/m3e, 5 mg/m3e [Dust]e, 1 mg/m3e [Fume]e NIOSH-REL: 1 mg/m3d Mercury 7439-97-6 3x10-4* 3x10-4* Dk None OSHA-PEL: 0.05 mg/m3g, 0.01 mg/m3 [Alkyl. cpds.]e, 0.01 (mercury, mg/m3 [Aryl. cpds.]e inorganic) NIOSH-REL: 0.05 mg/m3 [8 hr TWA] [skin]d Methane 74-82-8 NA NA NA NA Simple asphyxianth Methanol 67-56-1 5x10-1 1.75c NA None OSHA-PEL: 260 mg/m3, 325 mg/m3 [STEL] [skin]g ACGIH-TLV: 262 mg/m3e, 328 mg/m3 [STEL]e NIOSH-REL: 260 mg/m3 [skin], 325 mg/m3 [ceiling] [skin]f Methyl bromide 74-83-9 0.133b 0.465c NA None OSHA-PEL: 20 mg/m3g ACGIH-TLV: 19 mg/m3 (SKIN)e NIOSH-REL: Reduce to lowest level, the IDLH is 7900 mg/m3 Methylene chloride 75-09-2 6x10-2* 0.21c B2 7.5x10-3 OSHA-PEL: 1,765 mg/m3 [8-hr TWA], 3500 mg/m3 [ceiling]d (oral) ACGIH-TLV: 174 mg/m3e NIOSH-REL: reduce exposure to lowest feasible concentrationd Methyl ethyl ketone 78-93-3 0.6* 0.21c D None OSHA-PEL: 590 mg/m3 [8-hr TWA], 885 mg/m3 [15-min ST (2-butanone) limit]d ACGIH-TLV: same as aboved NIOSH-REL: same as abovei Methyl isobutyl ketone 108-10-1 0.05* 0.175c NA None OSHA-PEL: 205 mg/m3, 300 mg/m3 [STEL]d ACGIH-TLV: 205 mg/m3, 300 mg/m3 [STEL]e NIOSH-REL: 205 mg/m3, 300 mg/m3 [STEL]f Methyl-tert-butyl ether 1634-04-4 0.143j 0.5k NA None OSHA-PEL: i ACGIH-TLV:i NIOSH-REL:i Naphtha 8030-30-6 8.4b 29.4c NA None OSHA-PEL: 400 mg/m3d ACGIH-TLV: 1,200 mg/m3h NIOSH-REL: 400 mg/m3d Naphthalene 91-21-3 4x10-3# 0.0139c D None OSHA-PEL: 50 mg/m3d, 75 mg/m3 [STEL]d ACGIH-TLV: 52 mg/m3e, 79 mg/m3 [STEL]e NIOSH-REL: 50 mg/m3d Nickel 7440-02-0 2x10-2 0.02 A 0.84 OSHA-PEL: 0.1 mg/m3d ACGIH-TLV: 1 mg/m3e NIOSH-REL: .015 mg/m3d Nickel oxide 7440-02-0 3x10-4 1.2x10-3 NA None OSHA-PEL: 0.1 mg/m3d ACGIH-TLV: 0.1 mg/m3e Nitric acid 7697-37-2 0.035b 0.123c NA None OSHA-PEL: 5 mg/m3 and a 15 minute STEL 10 mg/m3d ACGIH-TLV: 5.2 mg/m3e, 10mg/m3 [STEL]e NIOSH-REL: same as aboved Nitrocellulose 9004-70-0 2.0x10-1 m 7.0x10-1 c e None OSHA-PEL: NA (cellulose tetranitrate) ACGIH-TLV: NA RTECS-LD50: >5g/kg (oral rat/mouse)5 Perfluoroalkylether 26675-46-7 1.3h 4.6m e None OSHA-PEL: NA (isoflurane) ACGIH-TLV: NA Sax/Japan TWA LC50: Inhal. Rat - 115,454 mg/m3h Perfluoro compounds 76-14-2 49j 171.3 e None OSHA-PEL: 7,000 mg/m3 (e.g., tetra CFCs) ACGIH-TLV: 6,990 mg/m3 NIOSH-REL: 7,000 mg/m3 Phenol 108-95-2 0.6* 2.098c Dg None OSHA-PEL: 19 mg/m3d ACGIH-TLV: 19 mg/m3e NIOSH-REL: 19 mg/m3d Phosphoric acid 7664-38-2 7x10-3 2.45x10-2 c NA None OSHA-PEL: 1 mg/m3d ACGIH-TLV: 1 mg/m3e NIOSH-REL: 1 mg/m3d and an IDLH of 10,000 mg/m3d Propane 74-98-6 12.6l 44.1c NA None OSHA-PEL: 1,800 mg/m3 ACGIH-TLV: 1,800 mg/m3 NIOSH-REL: 1,800 mg/m3 Propylene dichloride 78-87-5 2.429b 8.494c B2 0.068 OSHA-PEL: 350 mg/m3 and a STEL of 510 mg/m3d ACGIH-TLV: 347 mg/m3e, 508 mg/m3 [STEL]e NIOSH-IDLH: 9,400 mg/m3d Propyl glycol methyl ether 107-98-2 0.572j 2k NA None OSHA-PEL: NA (propylene glycol monomethyl ACGIH-TLV: 369 mg/m3 ether) OSHA-TWA: 360 mg/m3 Resins and formers 584-84-9 2.8x10-4 l 9.8x10-4 c NA None OSHA-PEL: 0.04 mg/m3 [8hr. TWA] (toluene-2,4-diisocyanate) ACGIH-TLV: 0.036 mg/m3 NIOSH-REL: 0.04 mg/m3 Sodium hydroxide 1310-73-2 1.4x10-2 b 4.89x10-2 c NA None OSHA-PEL: 2 mg/m3d ACGIH-TLV: 2 mg/m3e NIOSH-REL: 2 mg/m3d Sodium hypochlorite 7681-52-9 i i i None OSHA-PEL: i ACGIH-TLV: i NIOSH-REL: i Sulfuric acid 7664-93-9 0.07l 0.245c NA None OSHA-PEL: 1 mg/m3g ACGIH-TLV: 1 mg/m3e, 3 mg/m3 [STEL]e NIOSH-REL: same as abovef Tetrachloroethane (1,1,2,2) 79-34-5 0.049l 0.172c Ck 0.2* OSHA-PEL: 7 mg/m3d ACGIH-TLV: 6.9 mg/m3e NIOSH-REL: 7 mg/m3d, the IDLH is 1,050 mg/m3d Tetrahydrofuran 109-99-9 4.13l 14.455 c NA None OSHA-PEL: 590 mg/m3 ACGIH-TLV: 590 mg/m3, 737 mg/m3 [STEL]e NIOSH-REL: 590 mg/m3 Toluene 108-88-3 0.2k 0.4k D None OSHA-PEL: 375 mg/m3, 560 mg/m3 [ST]d ACGIH-TLV: 188 mg/m3 [TWA]e NIOSH-REL: 375 mg/m3, 560 mg/m3 [STEL]f Trichloroethane (1,1,1) 71-55-6 3.5x10-2# 1.0* D None OSHA-PEL: 1,900 mg/m3d ACGIH-TLV: 1,910 mg/m3e Trichloroethane (1,1,2) 79-00-5 4x10-2* under C 5.7x10-2 OSHA-PEL: 45 mg/m3 (skin)d review (inhal)* ACGIH-TLV: 55 mg/m3e 1.4x10-2 NIOSH-REL: 45 mg/m3 (skin)d Trichloroethylene 79-01-6 7.35x10-3# 0.013 B2 6.0x10-3 OSHA-PEL: 270 mg/m3 STEL 1080 mg/m3d (PNL) (inhal) ACGIH-TLV: 270 mg/m3 and STEL of 1,070 mg/m3e 1.1x10-2 NIOSH-REL: 1,370 mg/m3h (oral) Uranium (235, 238) 7440-61-1 0.003 0.0105c A 2.4x10- OSHA-PEL: soluble cmpds-.05 mg/m3, insoluble cmpds-0.2 8/pCi mg/m3d (inhal) ACGIH-TLV: .2 mg/m3e NIOSH-REL: soluble .2 mg/m3, insoluble .05 mg/m3d Vinyl acetate 108-05-4 0.057c 0.2k D None OSHA-PEL: 30 mg/m3, STEL 60 mg/m3f ACGIH-TLV: 35 mg/m3, STEL 70 mg/m3e NIOSH-REL: 15 mg/m3f Vinyl chloride 75-01-4 1.82x10-2 l 6.37x10-2 c A 0.3 OSHA-PEL: 2.6 mg/m3d (inhal) ACGIH-TLV: 13 mg/m3e Xylene 1330-20-7 2* 7c D None OSHA-PEL: 435 mg/m3, 655 mg/m3 [STEL]g ACGIH-TLV: 434 mg/m3e NIOSH-REL: 435 mg/m3,655 mg/m3d a The majority of the Reference Dose and Reference Concentration values in this table were taken from the Integrated Risk Information System (EPA 1993c) for the particular chemical. In a few cases, values were copied from the Health Effects Assessment Summary Tables. Values from the HEAST tables are denoted with the symbol (*) (EPA 1993a). Values followed by the symbol (#) were taken from the Office of Drinking Water's Health Advisories. Values from Pacific Northwest Laboratories of Battelle 1991 are denoted as (PNL). The Cancer Class and Slope Factor are, likewise, from the Integrated Risk Information System unless indicated otherwise (EPA 1993c). b Reference Dose calculated from the ACGIH-TLV, formula from the Center for Risk Management, Oak Ridge National Lab (1992d) (DOE 1992). c Reference Concentration calculated from the Reference Dose, using formula from the Center for Risk Management, Oak Ridge National Lab (DOE. 1992). d NIOSH 1990a. e ACGIH 1992a. f DHHS 1992b. g 29 CFR 1910. h Lewis 1992a. i Information is not available. j Reference Dose calculated from the value for the Reference Concentration. k ORNL 1993b. l Reference Dose calculated from the OSHA PEL/TWA. m Reference Dose calculated from the RTECS LD50. p Reference Concentration calculated from the NIOSH REL. r Calculated from the PNL RFD value (3.0) using the ORNL method. s Registry of toxic effects of chemical substances, DHHS, NIOSH, 1982. Note: NA - not applicable. E.3.4 Hazardous Chemical Risk/Effects Calculations Tables E.3.4-1 through E.3.4-36 present the calculations of HIs and cancer risks for each option for public exposure and workers at all 5 sites. The table numbers correspond exactly to the summary tables appearing in the text, with the risks for all options appearing in the summary table. Table E.3.4-1.-Risk Assessments from Exposure to Hazardous Chemicals from No Action Operation at Idaho National Engineering Laboratory Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetone 10.500 1,800 1,800 0 8.6x10-8 9.3x10-4 8.2x10-9 5.1x10-7 0 0 Ammonia 0.100 27 17 0 1.1x10-5 0.12 1.1x10-4 4.3x10-3 0 0 Nitric Acid 0.123 5 5 0 7.9x10-6 0.09 6.4x10-5 0.017 0 0 1,1,1-Trichloroethane 1.000 1,900 1,900 0 3.6x10-6 0.04 3.6x10-6 2.0x10-5 0 0 (TCA) Hazard Index - - - - - - 1.7x10-4 0.021 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-2.-Risk Assessments from Exposure to Hazardous Chemicals from Heavy Water Reactor Operation at Idaho National Engineering Laboratory Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 2.3x10-6 0.025 3.4x10-8 9.0x10-6 0 0 Ethanol 46.1 1,900 1,880 None 8.5x10-7 9.2x10-3 1.8x10-8 4.8x10-6 0 0 Methane (simple None None None None 2.3x10-6 0.03 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 8.5x10-7 9.2x10-3 4.8x10-7 3.5x10-5 0 0 Nitric acid 0.123 5 5 None 4.3x10-6 0.046 3.5x10-5 9.3x10-3 0 0 1,1,1-Trichloroethane 1.0 1,900 1,900 None 7.1x10-7 7.7x10-3 7.1x10-7 4.1x10-6 0 0 (TCA) Trichlorotrifluoroethane 105.0 7,600 7,600 None 3.0x10-5 0.33 2.9x10-7 4.3x10-5 0 0 (Freon 113) Hazard Index - - - - - - 3.6x10-5 9.4x10-3 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-3.-Risk Assessments from Exposure to Hazardous Chemicals from Modular High Temperature Gas-Cooled Reactor Operation at Idaho National Engineering Laboratory Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 2.3x10-6 0.03 3.4x10-8 9.0x10-6 0 0 Ethanol 46.1 1,900 1,880 None 8.5x10-7 9.2x10-3 1.8x10-8 4.8x10-6 0 0 Methane (simple None None None None 2.3x10-6 0.03 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 8.5x10-7 9.2x10-3 4.8x10-7 3.5x10-5 0 0 1,1,1-Trichloroethane 1.0 1,900 1,900 None 2.3x10-7 0.03 2.3x10-7 1.3x10-5 0 0 (TCA) Hazard Index - - - - - - 7.7x10-7 6.2x10-5 0 - Total Cancer Risk - - - - - - 0 0 0 0 Table E.3.4-4.-Risk Assessments from Exposure to Hazardous Chemicals from Advanced Light Water Reactor Operation at Idaho National Engineering Laboratory Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetone 10.5 1,800 1,800 None 7.1x10-6 0.08 6.8x10-7 4.3x10-5 0 0 Acetylene 66.28 2,708 2,708 None 2.3x10-6 0.03 3.4x10-8 9.0x10-6 0 0 Ammonia 0.1 27 17 None 3.7x10-6 0.04 3.7x10-5 1.5x10-3 0 0 Ethanol 46.1 1,900 1,880 None 8.5x10-7 9.2x10-3 1.8x10-8 4.8x10-6 0 0 Methane (simple None None None None 2.3x10-6 0.03 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 8.5x10-7 9.2x10-3 4.8x10-7 3.5x10-5 0 0 Nitric acid 0.123 5 5 None 5.0x10-5 0.54 4.1x10-4 0.11 0 0 1,1,1-Trichloroethane 1.0 1,900 1,900 None 9.8x10-6 0.11 9.8x10-6 5.6x10-6 0 0 (TCA) Hazard Index - - - - - - 4.5x10-4 0.11 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-5.-Risk Assessments from Exposure to Hazardous Chemicals from Accelerator Production of Tritium Operation at Idaho National Engineering Laboratory Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concen-tration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 2.3x10-6 0.03 3.4x10-8 9.0x10-6 0 0 Ethanol 46.1 1,900 1,880 None 8.5x10-7 9.2x10-3 1.8x10-8 4.8x10-6 0 0 Methane (simple None None None None 2.3x10-6 0.03 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 8.5x10-7 9.2x10-3 4.8x10-7 3.5x10-5 0 0 Hazard Index - - - - - - 5.4x10-7 4.9x10-5 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-6.-Risk Assessments from Exposure to Hazardous Chemicals from Tritium Recycling Operation at Idaho National Engineering Laboratory Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 2.3x10-6 0.03 3.4x10-8 9.0x10-6 0 0 Ethanol 46.1 1,900 1,880 None 8.5x10-7 9.2x10-3 1.8x10-8 4.8x10-6 0 0 Methane (simple None None None None 2.3x10-6 0.03 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 8.5x10-7 9.2x10-3 4.8x10-7 3.5x10-5 0 0 Hazard Index - - - - - - 5.4x10-7 4.9x10-5 - - Total Cancer Risk - - - - - - - - - 0 Table E.3.4-7.-Risk Assessments from Exposure to Hazardous Chemicals at Idaho National Engineering Laboratory-Summary Hazard Index and Total Cancer Risk Technology Hazard Index Total Cancer Risk - Boundary Worker 100 Boundary Worker 100 MEI, Meters MEI, Meters Annual 8 Hoursb Annual 8 Hoursd, No Action 1.7x10-4 0.021 0 0 Tritium Recycle 5.4x10-7 4.9x10-5 0 0 Heavy Water Reactor 3.6x10-5 9.4x10-3 0 0 Heavy Water Reactor and Tritum Recycling 3.7x10-5 9.4x10-3 0 0 Heavy Water Reactor, Tritum Recycling, and No Action 2.1x10-4 0.031 0 0 Modular High Temperature Gas-Cooled Reactor 7.7x10-7 6.2x10-5 0 0 Modular High Temperature Gas-Cooled Reactor and Tritium 1.3x10-6 1.1x10-4 0 0 Recycling Modular High Temperature Gas-Cooled Reactor, Tritium 1.8x10-4 0.021 0 0 Recycling, and No Action Advanced Light Water Reactor 4.5x10-4 0.11 0 0 Advanced Light Water Reactor and Tritium Recycling 4.6x10-4 0.11 0 0 Advanced Light Water Reactor, Tritium Recycling, and No 6.3x10-4 0.13 0 0 Action Accelerator Production of Tritium 5.4x10-7 4.9x10-5 0 0 Accelerator Production of Tritium and Tritium Recycling 5.4x10-6 9.8x10-5 0 0 Accelerator Production of Tritium, Tritium Recycling, and No 1.8x10-4 0.021 0 0 Action Table E.3.4-8.-Risk Assessments from Exposure to Hazardous Chemicals from No Action Operation at Nevada Test Site Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) None None None None None None None None None None None Hazard Index - - - - - - None None None None Total Cancer Risk - - - - - - None None None None Table E.3.4-9.-Risk Assessments from Exposure to Hazardous Chemicals from Heavy Water Reactor Operation at Nevada Test Site Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 3.9x10-7 8.4x10-3 5.8x10-9 3.1x10-6 0 0 Ethanol 46.1 1,900 1,880 None 1.4x10-7 3.2x10-3 3.1x10-9 1.7x10-6 0 0 Methane (simple None None None None 3.9x10-7 8.4x10-3 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 1.4x10-7 3.2x10-3 8.2x10-8 1.2x10-5 0 0 Nitric acid 0.123 5 5 None 7.3x10-7 0.016 5.9x10-6 3.2x10-3 0 0 1,1,1-Trichloroethane 1.0 1,900 1,910 None 1.2x10-7 2.7x10-3 1.2x10-7 1.4x10-6 0 0 (TCA) Trichlorotrifluoroethane 105.0 7,600 7,670 None 5.2x10-6 0.11 4.9x10-8 1.5x10-5 0 0 (Freon 113) Hazard Index - - - - - - 6.2x10-6 3.2x10-3 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-10.-Risk Assessments from Exposure to Hazardous Chemicals Modular High Temperature Gas-Cooled Reactor Operation at Nevada Test Site Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 3.9x10-7 8.4x10-3 5.8x10-9 3.1x10-6 0 0 Ethanol 46.1 1,900 1,880 None 1.4x10-7 3.2x10-3 3.1x10-9 1.7x10-6 0 0 Methane (simple None None None None 3.9x10-7 8.4x10-3 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 1.4x10-7 3.2x10-3 8.2x10-8 1.2x10-5 0 0 1,1,1-Trichloroethane 1.0 1,900 1,910 None 3.9x10-8 8.6x10-4 3.9x10-8 4.5x10-7 0 0 (TCA) Hazard Index - - - - - - 1.3x10-7 1.7x10-5 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-11.-Risk Assessments from Exposure to Hazardous Chemicals from Advanced Light Water Reactor Operation at Nevada Test Site Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetone 10.5 1,800 1,800 None 1.2x10-6 0.03 1.2x10-7 1.5x10-5 0 0 Acetylene 66.28 2,708 2,708 None 3.9x10-7 8.4x10-3 5.8x10-9 3.1x10-6 0 0 Ammonia 0.1 27 17 None 6.4x10-7 0.014 6.4x10-6 5.1x10-4 0 0 Ethanol 46.1 1,900 1,880 None 1.4x10-7 3.2x10-3 3.1x10-9 1.7x10-6 0 0 Methane (simple None None None None 3.9x10-7 8.4x10-3 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 1.4x10-7 3.2x10-3 8.2x10-8 1.2x10-5 0 0 Nitric acid 0.123 5 5 None 8.5x10-6 0.19 6.9x10-5 0.037 0 0 1,1,1-Trichloroethane 1 1,900 1,900 None 1.7x10-6 3.7x10-2 1.7x10-6 1.9x10-5 0 0 (TCA) Hazard Index - - - - - - 7.7x10-5 3.8x10-2 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-12.-Risk Assessments from Exposure to Hazardous Chemicals from Accelerator Production of Tritium Operation at Nevada Test Site Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 3.9x10-7 8.4x10-3 5.8x10-9 3.1x10-6 0 0 Ethanol 46.1 1,900 1,880 None 1.4x10-7 3.2x10-3 3.1x10-9 1.7x10-6 0 0 Methane (simple None None None None 3.9x10-7 8.4x10-3 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 1.4x10-7 3.2x10-3 8.2x10-8 1.2x10-5 0 0 Hazard Index - - - - - - 9.1x10-8 1.7x10-5 0 0 Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-13.-Risk Assessments from Exposure to Hazardous Chemicals from Tritium Recycling Operation at Nevada Test Site Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 3.9x10-7 8.4x10-3 5.8x10-9 3.1x10-6 0 0 Ethanol 46.1 1,900 1,880 None 1.4x10-7 3.2x10-3 3.1x10-9 1.7x10-6 0 0 Methane (simple None None None None 3.9x10-7 8.4x10-3 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 1.4x10-7 3.2x10-3 8.2x10-8 1.2x10-5 0 0 Hazard Index - - - - - - 9.1x10-8 1.7x10-5 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-14.-Risk Assessments from Exposure to Hazardous Chemicals at Nevada Test Site- Summary Hazard Index and Total Cancer Risk Technology Hazard Index Total Cancer Risk - Boundary Worker 100 Boundary Worker 100 MEI, Meters MEI, Meters Annual 8 Hoursb Annual 8 Hoursd, No Action 0 0 0 0 Tritium Recycle 9.1x10-8 1.7x10-5 0 0 Heavy Water Reactor 6.2x10-6 3.2x10-3 0 0 Heavy Water Reactor and Tritum Recycling 6.3x10-6 3.2x10-3 0 0 Heavy Water Reactor, Tritum Recycling, and No Action 6.3x10-6 3.2x10-3 0 0 Modular High Temperature Gas-Cooled Reactor 1.3x10-7 1.7x10-5 0 0 Modular High Temperature Gas-Cooled Reactor and 2.2x10-7 3.4x10-5 0 0 Tritium Recycling Modular High Temperature Gas-Cooled Reactor, 2.2x10-7 3.4x10-5 0 0 Tritium Recycling and No Action Advanced Light Water Reactor 7.7x10-5 0.038 0 0 Advanced Light Water Reactor and Tritium Recycling 7.7x10-5 0.038 0 0 Advanced Light Water Reactor, Tritium Recycling, and 7.7x10-5 0.038 0 0 No Action Accelerator Production of Tritium 9.1x10-8 1.7x10-5 0 0 Accelerator Production of Tritium and Tritium Recycling 1.8x10-7 3.4x10-5 0 0 Accelerator Production of Tritium, Tritium Recycling, and 1.8x10-7 3.4x10-5 0 0 No Action Table E.3.4-15.-Risk Assessments from Exposure to Hazardous Chemicals from No Action Operation at Oak Ridge Reservation Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Chlorine 0.004 1.5 1.5 None 3.0x10-4 0.06 0.08 0.04 0 0 Chlorodifluoromethane 86.7 3,540 3,540 None 6.4x10-4 0.12 7.3x10-6 3.4x10-5 0 0 (Freon 22) Dichlorodifluoromethane 0.7 4,950 4,950 None 5.6x10-4 0.11 8.0x10-4 2.2x10-5 0 0 (Freon 12) Hydrogen chloride 0.007 7 7 None 1.3x10-3 0.25 0.18 0.04 0 0 Methanol 1.75 260 200 None 4.8x10-3 0.92 2.7x10-3 3.6x10-3 0 0 Nitric acid 0.123 5 5 None 1.7x10-3 0.33 0.01 0.07 0 0 Sulfuric acid 0.025 1 1 None 4.5x10-4 0.09 0.02 0.09 0 0 Tetrachloroethylene 0.035 170 None None 2.2x10-3 0.43 0.06 2.5x10-3 0 0 (PERC) 1,1,1-Trichloroethane 1 1 1,900 None 1.4x10-4 0.03 1.4x10-4 0.03 0 0 (TCA) Trichlorotrifluoroethane 105 7,600 7,670 None 5.0x10-4 0.10 4.8x10-6 1.3x10-5 0 0 (Freon 113) Trichlorofluoromethane 1.05 5,620 5,620 None 1.5x10-3 0.28 1.4x10-3 5.0x10-5 0 0 (Freon 11) Hazard Index - - - - - - 0.36 0.26 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-16.-Risk Assessments from Exposure to Hazardous Chemicals from Heavy Water Reactor Operation at Oak Ridge Reservation Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 1.1x10-4 0.02 1.6x10-6 7.4x10-6 0 0 Ethanol 46.1 1,900 1,880 None 3.9x10-5 7.5x10-3 8.5x10-7 4.0x10-6 0 0 Methane (simple None None None None 1.1x10-4 0.02 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 3.9x10-5 7.5x10-3 2.2x10-5 2.9x10-5 0 0 Nitric acid 0.12 5 5 None 2.0x10-4 0.04 1.6x10-3 7.6x10-3 0 0 1,1,1-Trichloroethane 1 1,900 1,910 None 3.3x10-5 6.3x10-3 3.3x10-5 3.3x10-6 0 0 (TCA) Trichlorotrifluoroethane 105 7,600 7,670 None 1.4x10-3 0.27 1.3x10-5 3.5x10-5 0 0 (Freon 113) Hazard Index - - - - - - 1.7x10-3 7.7x10-3 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-17.-Risk Assessments from Exposure to Hazardous Chemicals from Modular High Temperature Gas-Cooled Reactor Operation at Oak Ridge Reservation Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 1.1x10-4 0.02 9.6x10-5 0.02 0 0 Ethanol 46.1 1,900 1,880 None 3.9x10-5 7.5x10-3 4.8x10-5 9.1x10-3 0 0 Methane (simple None None None None 1.1x10-4 0.02 9.6x10-5 0.02 0 0 asphyxiant) Methanol 1.75 260 200 None 3.9x10-5 7.5x10-3 4.8x10-5 9.1x10-3 0 0 1,1,1-Trichloroethane 1 1,900 1,910 None 1.1x10-5 2.1x10-3 1.1x10-5 2.1x10-3 0 0 (TCA) Hazard Index - - - - - - 3.0x10-4 0.057 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-18.-Risk Assessments from Exposure to Hazardous Chemicals from Advanced Light Water Reactor Operation at Oak Ridge Reservation Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetone 10.5 1,800 1,800 None 3.3x10-4 0.06 3.2x10-5 3.5x10-5 0 0 Acetylene 66.28 2,708 2,708 None 1.1x10-4 0.02 1.6x10-6 7.4x10-6 0 0 Ammonia 0.1 27 17 None 1.7x10-4 0.03 1.7x10-3 1.2x10-3 0 0 Ethanol 46.1 1,900 1,880 None 3.9x10-5 7.5x10-3 8.5x10-7 4.0x10-6 0 0 Methane (simple None None None None 1.1x10-4 0.02 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 3.9x10-5 7.5x10-3 2.2x10-5 2.9x10-5 0 0 Nitric acid 0.12 5 5 None 2.3x10-3 0.44 0.019 0.09 0 0 1,1,1-Trichloroethane 1 1,900 1,900 None 4.6x10-4 8.7x10-2 4.6x10-4 4.6x10-5 0 0 (TCA) Hazard Index - - - - - - 0.02 0.09 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-19.-Risk Assessments from Exposure to Hazardous Chemicals from Accelerator Production of Tritium Operation at Oak Ridge Reservation Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 1.1x10-4 0.02 1.6x10-6 7.4x10-6 0 0 Ethanol 46.1 1,900 1,880 None 3.9x10-5 7.5x10-3 8.5x10-7 4.0x10-6 0 0 Methane (simple None None None None 1.1x10-4 0.02 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 3.9x10-5 7.5x10-3 2.2x10-5 2.9x10-5 0 0 Hazard Index - - - - - - 2.5x10-5 4.0x10-5 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-20.-Risk Assessments from Exposure to Hazardous Chemicals from Tritium Recycling Operation at Oak Ridge Reservation Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 1.1x10-4 0.02 1.6x10-6 7.4x10-6 0 0 Ethanol 46.1 1,900 1,880 None 3.9x10-5 7.5x10-3 8.5x10-7 4.0x10-6 0 0 Methane (simple None None None None 1.1x10-4 0.02 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 3.9x10-5 7.5x10-3 2.2x10-5 2.9x10-5 0 0 Hazard Index - - - - - - 2.5x10-5 4.0x10-5 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-21.-Risk Assessments from Exposure to Hazardous Chemicals at Oak Ridge Reservation-Summary Hazard Index and Total Cancer Risk Technology Hazard Index Total Cancer Risk - Boundary Worker 100 Boundary Worker 100 MEI, Meters MEI, Meters Annual 8 Hoursb Annual 8 Hoursd, No Action 0.36 0.26 0 0 Tritium Recycle 2.5x10-5 4.0x10-5 0 0 Heavy Water Reactor 1.7x10-3 7.7x10-3 0 0 Heavy Water Reactor and Tritium Recycling 1.7x10-3 7.7x10-3 0 0 Heavy Water Reactor, Tritium Recycling and No Action 0.36 0.27 0 0 Modular High Temperature Gas-Cooled Reactor 3.0x10-4 0.06 0 0 Modular High Temperature Gas-Cooled Reactor and Tritium Recycling 3.2x10-4 0.06 0 0 Modular High Temperature Gas-Cooled Reactor, Tritium Recycling and No 0.36 0.32 0 0 Action Advanced Light Water Reactor 0.021 0.090 0 0 Advanced Light Water Reactor and Tritium Recycling 0.022 0.090 0 0 Advanced Light Water Reactor, Tritium Recycling and No Action 0.38 0.35 0 0 Accelerated Production of Tritium 2.5x10-5 4.0x10-5 0 0 Accelerated Production of Tritium and Tritium Recycling 5.0x10-5 8.1x10-5 0 0 Accelerated Production of Tritium, Tritium Recycling and No Action 0.36 0.26 0 0 Table E.3.4-22.-Risk Assessments from Exposure to Hazardous Chemicals from No Action Operation at Pantex Plant [Page 1 of 3] Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetone 10.5 1,800.0 1,800.0 None 2.0x10-4 0.39 1.9x10-5 2.2x10-4 0 0 Aliphatic alcohols 0.35 150.0 - None 7.6x10-8 5.5x10-4 2.2x10-7 3.7x10-6 0 0 (n-butyl alcohol) Aliphatic hydrocarbons 25.7 1,050.0 1,050.0 None 3.2x10-6 0.02 1.2x10-7 2.0x10-5 0 0 (cyclohexane) 2-butoxyethanol 2.96 240.0 - None 1.3x10-4 0.24 4.4x10-5 9.9x10-4 0 0 Butyl acetate 17.4 710.0 710.0 None 1.4x10-5 0.10 8.0x10-7 1.4x10-4 0 0 Butyl alcohol 0.35 150.0 152.0 None 5.2x10-6 0.04 1.5x10-5 2.4x10-4 0 0 Chlorodifluoromethane 86.7 3,540.0 - None 3.0x10-6 5.5x10-3 3.5x10-8 1.6x10-6 0 0 (Freon 142) Cyanogen 0.245 20.0 - None 7.6x10-6 0.18 3.1x10-5 9.2x10-3 0 0 Cyclohexane 25.7 1,050.0 1,050.0 None 1.8x10-6 3.3x10-3 7.1x10-8 3.2x10-6 0 0 Diacetone alcohol 5.88 240.0 - None 1.5x10-7 1.1x10-3 2.6x10-8 4.6x10-6 0 0 Dibutyl phthalate 0.35 5.0 5.0 None 2.3x10-7 1.7x10-3 6.5x10-7 3.3x10-4 0 0 Dichlorodifluromethane 0.7 4,950.0 4,950.0 None 4.3x10-6 0.03 6.2x10-6 5.4x10-6 0 0 (Freon 12) Diesel (undefined - - - None 1.5x10-6 2.8x10-3 0 0 0 0 complex) Diethylene glycol ethyl 0.35 50.0 - None 7.6x10-8 5.5x10-4 2.2x10-7 1.1x10-5 0 0 ether Epoxy solvent (toluene) 0.4 375.0 - None 2.6x10-6 0.02 6.5x10-6 5.0x10-5 0 0 Ethanol 46.1 1,900.0 1,880.0 None 3.5x10-6 0.01 7.6x10-8 7.3x10-6 0 0 Ethyl acetate 3.15 1,400.0 1,400.0 None 9.9x10-7 7.2x10-3 3.1x10-7 5.1x10-6 0 0 Ethyl benzene 1.0 435.0 - None 1.5x10-7 1.1x10-3 1.5x10-7 2.5x10-6 0 0 Freon TF (Freon 11) 1.05 5,600.0 5,620.0 None 1.6x10-5 0.03 1.5x10-5 5.0x10-6 0 0 Heptane 40.1 2,000.0 1,640.0 None 3.8x10-7 2.8x10-3 9.5x10-9 1.4x10-6 0 0 Hydrocarbons (heptane) 40.1 2,000.0 - None 8.0x10-5 2.06 2.0x10-6 1.0x10-3 0 0 Hydrogen chloride 0.007 7.0 7.0 None 1.1x10-5 0.26 1.6x10-3 0.04 0 0 Hydrogen cyanide 0.07 5.0 - None 1.1x10-6 0.03 1.5x10-5 5.1x10-3 0 0 Hydrogen fluoride 0.063 2.5 2.6 None 1.3x10-5 0.33 2.1x10-4 0.13 0 0 Isopropyl alcohol 24.15 980.0 980.0 None 2.2x10-5 0.14 8.9x10-7 1.5x10-4 0 0 Isobutane 35.04 1,430.0 - None 2.3x10-7 1.7x10-3 6.5x10-9 1.2x10-6 0 0 Isobutyl acetate 17.15 700.0 - None 3.8x10-7 2.8x10-3 2.2x10-8 3.9x10-6 0 0 Methanol 1.75 260.0 200.0 None 1.1x10-4 0.20 6.3x10-5 7.8x10-4 0 0 Methyl ethyl ketone 1.00 590.0 590.0 None 1.5x10-5 0.11 1.5x10-5 1.8x10-4 0 0 Methyl isobutyl ketone 0.05 205.0 205.0 None 1.7x10-6 6.1x10-3 3.5x10-5 3.0x10-5 0 0 Methylene chloride 3.00 1,765.0 - 0.0075 7.6x10-7 2.2x10-3 2.5x10-7 1.2x10-6 1.6x10-9 6.4x10-7 Nitrocellulose 0.7 100.0 - None 5.3x10-7 3.9x10-3 7.6x10-7 3.9x10-5 0 0 Perfluoroalkylether 17.76 2,537.143 - None 7.6x10-8 5.5x10-4 4.3x10-9 2.2x10-7 0 0 (isoflurane) Perfluoro compounds 171.3 7,000.0 - None 3.0x10-7 5.5x10-4 1.8x10-9 7.9x10-8 0 0 (Freon 114) Propane 44.1 1,800.0 - None 6.8x10-7 5.0x10-3 1.6x10-8 2.8x10-6 0 0 Propyl glycol methyl 9.04 369.0 - None 3.0x10-7 2.2x10-3 3.4x10-8 6.0x10-6 0 0 ether (propylene glycol monomethyl ether) Resins and formers 0.001 0.14 - None 1.3x10-6 9.4x10-3 1.3x10-3 0.07 0 0 (toluene-2,4 diisocyanate) Methyl tert butyl ether 0.5 71.429 - None 5.1x10-6 2.2x10-3 1.0x10-5 3.1x10-5 0 0 Tetrahydrofuran 14.46 590.0 - None 1.9x10-5 0.03 1.3x10-6 5.7x10-5 0 0 Toluene 0.4 375.0 383.0 None 1.3x10-4 0.27 3.2x10-4 7.1x10-4 0 0 1,1,1 Trichloroethane 1.0 1,900.0 1,900.0 None 2.2x10-6 9.4x10-3 2.2x10-6 4.9x10-6 0 0 (TCA) Trichloroethylene 0.046 270.0 - 0.006 7.9x10-8 5.5x10-4 1.7x10-6 2.0x10-6 1.4x10-10 1.3x10-7 Trichlorofluoromethane 1.05 5,600.0 5,620.0 None 4.6x10-7 1.7x10-3 4.3x10-7 2.9x10-7 0 0 (Freon 11) 1,1,2-Trichloro- 105.0 7,600.0 7,600.0 None 3.3x10-6 0.01 3.2x10-8 1.9x10-6 0 0 1,2,2-trifluoroethane (Freon 113) Trichlorotrifluoroethane 105.0 7,600.0 7,600.0 None 4.1x10-6 8.3x10-3 3.9x10-8 1.1x10-6 0 0 (Freon 113) VM and P naptha 29.4 400.0 - None 3.9x10-6 0.03 1.3x10-7 6.6x10-5 0 0 Xylene 7.0 435.0 - None 2.7x10-6 0.05 3.9x10-7 1.2x10-4 - - Hazard Index - - - - - - 3.7x10-3 2.6x10-1 - - Total Cancer Risk - - - - - - - - 1.3x10-9 7.7x10-7 Table E.3.4-23.-Risk Assessments from Exposure to Hazardous Chemicals from Heavy Water Reactor Operation at Pantex Plant Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 1.9x10-3 8.1x10-3 2.8x10-5 3.0x10-6 0 0 Ethanol 46.10 1,900 1,880 None 6.9x10-6 3.0x10-3 1.5x10-7 1.6x10-6 0 0 Methane (simple None None None None 1.9x10-3 8.1x10-3 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 6.9x10-6 3.0x10-3 4.0x10-6 1.2x10-5 0 0 Nitric acid 0.123 5 5 None 3.5x10-5 0.02 2.8x10-4 3.0 x10-3 0 0 1,1,1-Trichloroethane 1.0 1,900 1,900 None 5.8x10-6 2.5x10-3 5.8x10-6 1.3x10-6 0 0 (TCA) Trichlorotrifluoroethane 105.0 7,600 7,600 None 2.5x10-4 0.11 2.4x10-6 1.4x10-5 0 0 (Freon 113) Hazard Index - - - - - - 3.2x10-4 3.1x10-3 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-24.-Risk Assessments from Exposure to Hazardous Chemicals from Modular High Temperature Gas-Cooled Reactor Operation at Pantex Plant Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 1.7x10-5 7.3x10-3 2.5x10-7 2.7x10-6 0 0 Ethanol 46.10 1,900 1,880 None 6.2x10-6 2.7x10-3 1.3x10-7 1.4x10-6 0 0 Methane (simple None None None None 1.7x10-5 7.3x10-3 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 6.2x10-6 2.7x10-3 3.5x10-6 1.0x10-5 0 0 1,1,1-Trichloroethane 1.00 1,900 1,900 None 1.9x10-6 8.2x10-4 1.9x10-6 4.3x10-7 0 0 (TCA) Hazard Index - - - - - - 5.8x10-6 1.5x10-5 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-25.-Risk Assessments from Exposure to Hazardous Chemicals from Advanced Light Water Reactor Operation at Pantex Plant Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetone 10.5 1,800 1,800 None 5.8x10-5 0.03 5.6x10-6 1.4x10-5 0 0 Acetylene 66.28 2,708 2,708 None 1.9x10-5 8.1x10-3 2.8x10-7 3.0x10-6 0 0 Ammonia 0.1 27 17 None 3.1x10-5 0.01 3.1x10-4 4.9x10-4 0 0 Ethanol 46.1 1,900 1,880 None 6.9x10-6 3.0x10-3 1.5x10-7 1.6x10-6 0 0 Methane (simple None None None None 1.9x10-5 8.1x10-3 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 6.9x10-6 3.0x10-3 4.0x10-6 1.2x10-5 0 0 Nitric acid 0.123 5 5 None 4.1x10-4 0.02 3.3x10-3 3.5x10-3 0 0 1,1,1-Trichloroethane 1.0 1,900 1,900 None 8.0x10-5 3.5x10-2 8.0x10-5 1.8x10-5 0 0 (TCA) Hazard Index - - - - - - 3.7x10-3 4.1x10-3 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-26.-Risk Assessments from Exposure to Hazardous Chemicals from Accelerator Production of Tritium Operation at Pantex Plant Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 1.9x10-3 8.1x10-3 2.8x10-5 3.0x10-6 0 0 Ethanol 46.10 1,900 1,880 None 6.9x10-6 3.0x10-3 1.5x10-7 1.6x10-6 0 0 Methane (simple None None None None 1.9x10-3 8.1x10-3 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 6.9x10-6 3.0x10-3 4.0x10-6 1.2x10-5 0 0 Hazard Index - - - - - - 3.2x10-5 1.6x10-5 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-27.-Risk Assessments from Exposure to Hazardous Chemicals from Tritium Recycling Operation at Pantex Plant Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 1.9x10-3 8.1x10-3 2.8x10-5 3.0x10-6 0 0 Ethanol 46.10 1,900 1,880 None 6.9x10-6 3.0x10-3 1.5x10-7 1.6x10-6 0 0 Methane (simple None None None None 1.9x10-3 8.1x10-3 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 6.9x10-6 3.0x10-3 4.0x10-6 1.2x10-5 0 0 Hazard Index - - - - - - 3.2x10-5 1.6x10-5 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-28.-Risk Assessments from Exposure to Hazardous Chemicals at Pantex Plant-Summary Hazard Index and Total Cancer Risk Technology Hazard Index Total Cancer Risk - Boundary Worker 100 Boundary Worker 100 MEI, Meters MEI, Meters Annual 8 Hoursb Annual 8 Hoursd, No Action 3.7x10-3 0.26 1.8x10-9 7.7x10-7 Tritium Recycle 3.2x10-5 1.6x10-5 0 0 Heavy Water Reactor 3.2x10-4 3.1x10-3 0 0 Heavy Water Reactor and Tritium Recycling 3.6x10-4 3.1x10-3 0 0 Heavy Water Reactor, Tritium Recycling, and No Action 4.1x10-3 0.26 1.8x10-9 7.7 x10-7 Modular High Temperature Gas-Cooled Reactor 5.8x10-6 1.5x10-5 0 0 Modular High Temperature Gas-Cooled Reactor and Tritium 3.8x10-5 3.1x10-5 0 0 Recycling Modular High Temperature Gas-Cooled Reactor, Tritium 3.7x10-3 0.26 1.8x10-9 7.7 x10-7 Recycling, and No Action Advanced Light Water Reactor 3.7x10-3 4.1x10-3 0 0 Advanced Light Water Reactor and Tritium Recycling 3.8 x10-3 4.1x10-3 0 0 Advanced Light Water Reactor, Tritium Recycling, and No Action 7.5x10-3 0.26 1.8x10-9 7.7 x10-7 Accelerated Production of Tritium 3.2x10-5 1.6x10-5 0 0 Accelerated Production of Tritium and Tritium Recycling 6.4 x10-5 3.2x10-5 0 0 Accelerated Production of Tritium, Tritium Recycling and No 3.8 x10-3 0.26 1.8x10-9 7.7 x10-7 Action Table E.3.4-29.-Risk Assessments from Exposure to Hazardous Chemicals from No Action Operation at Savannah River Site [Page 1 of 2] Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acrolein 0.00002 0.25 0.23 None 1.6x10-6 2.1x10-3 0.08 0.01 0 0 Acrylonitrile 0.002 4.42 4.3 0.24 1.6x10-6 2.1x10-3 7.8x10-4 4.8x10-4 1.1x10-7 2.0x10-5 Antimony 0.0000139 0.5 0.5 None 9.2x10-7 1.2x10-3 0.07 2.5x10-3 0 0 Benzene 0.783 3.25 32.0 0.029 2.3x10-3 3.1 2.9x10-3 0.95 1.9x10-5 3.5x10-3 Cadmium 0.0005 0.2 0.05 None 9.2x10-7 1.3x10-3 1.8x10-3 6.3x10-3 0 0 Cadmium oxide 0.0025 0.1 0.05 None 1.1x10-6 1.4x10-3 4.2x10-4 0.014 0 0 Chlorine 0.039 1.5 1.0 None 1.2x10-4 0.16 3.1x10-3 0.11 0 0 2,4-Dinitrololuene 0.0069 1.5 0.15 None 1.8x10-5 0.03 2.7x10-3 0.02 0 0 Dioctyl phthalate 0.07 5.0 5.0 0.014 2.1x10-6 2.9x10-3 3.1x10-5 5.8x10-4 8.6x10-9 1.6x10-6 Ethyl benzene 1.0 435.0 434.0 None 2.1x10-5 0.03 2.1x10-5 6.7x10-5 0 0 Ethylene glycol 0.2 740.0 18.0 None 8.1x10-6 0.01 4.1x10-5 1.5x10-5 0 0 Formic acid 6.994 9.0 9.4 None 2.9x10-5 0.04 4.2x10-6 4.4x10-3 0 0 Hexane 0.2 1,800.0 176.0 None 4.1x10-6 5.5x10-3 2.0x10-5 3.1x10-6 0 0 Hydrogen chloride 0.007 7.0 7.0 None 2.9x10-3 3.9 0.41 0.55 0 0 Hydrogen sulfide 0.0009 14.0 14.0 None 8.3x10-5 0.11 0.09 8.1x10-3 0 0 Manganese 0.005 1.0 2.0 None 6.3x10-6 8.5x10-3 1.3x10-3 8.5x10-3 0 0 Mercury 0.0003 0.05 0.05 None 3.2x10-6 4.3x10-3 0.01 0.09 0 0 Methanol 1.75 260.0 200.0 None 6.4x10-6 8.6x10-3 3.6x10-6 3.3x10-5 0 0 Methylene chloride 3.0 1,765.0 177.0 0.0075 2.2x10-5 0.03 7.3x10-6 1.7x10-5 4.7x10-8 8.6x10-6 Methyl ethyl ketone 1.0 590.0 590.0 None 8.1x10-5 0.11 8.1x10-5 1.9x10-4 0 0 (MEK) Methyl tert-butyl ether 0.2 20.0 - None 3.1x10-5 0.04 1.6x10-4 2.1x10-3 0 0 Methyl isobutyl ketone 0.05 205.0 205.0 None 4.2x10-5 0.06 8.4x10-4 2.8x10-4 0 0 (MIK) Nickel 0.02 1.0 1.0 0.84 4.3x10-6 5.8x10-3 2.2x10-4 5.8x10-3 1.0x10-6 1.9x10-4 Nickel oxide 0.0012 0.1 0.1 None 9.2x10-7 1.2x10-3 7.7x10-4 0.01 0 0 Nitric acid 0.123 5.0 5.0 None 4.8x10-5 0.06 3.9x10-4 0.01 0 0 Sodium hydroxide 0.0489 2.0 2.0 None 6.3x10-6 8.6x10-3 1.3x10-4 4.3x10-3 0 0 Sulfuric acid 0.0245 1.0 1.0 None 9.5x10-7 1.3x10-3 3.9x10-5 1.3x10-3 0 0 Tetrachloroethylene 0.03496503 170.0 339.0 None 5.3x10-4 0.72 1.5x10-2 4.2x10-3 0 0 (PERC) Toluene 0.4 375.0 377.0 None 2.8x10-5 0.04 7.1x10-5 1.0x10-4 0 0 1,1,1-Trichloroethane 1.0 1,900.0 1,900.0 None 2.2x10-5 0.03 2.2x10-5 1.6x10-5 0 0 (111-TCA) Trichloroethylene (TCE) 0.046 270.0 270.0 0.006 1.8x10-4 0.24 3.9x10-3 9.0x10-4 3.1x10-7 5.6x10-5 Trichloromethane 0.03497 240.0 9.78 0.081 5.2x10-4 0.70 1.5x10-2 2.9x10-3 1.2x10-5 2.2x10-3 (chloroform) 1,1,2-Trichloro- 105.0 7,600.0 7,670.0 None 1.6x10-6 2.1x10-3 1.5x10-8 2.8x10-7 0 0 1,2,2-Trifluoroethane (Freon 113) Xylene 7.0 435.0 435.0 None 3.8x10-4 0.51 5.4x10-5 1.2x10-3 0 0 Hazard Index - - - - - - 0.70 1.8 - - Total Cancer Risk - - - - - - - - 3.2x10-5 5.9x10-3 Table E.3.4-30.-Risk Assessments from Exposure to Hazardous Chemicals from Heavy Water Reactor Operation at Savannah River Site Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 1.1x10-5 0.01 1.6x10-7 5.3x10-6 0 0 Ethanol 46.1 1,900 1,900 None 3.9x10-6 5.3x10-3 8.5x10-8 2.8x10-6 0 0 Methane (simple None None None None 1.1x10-5 0.01 0 0 0 0 asphyxiant) Methanol 1.75 260 260 None 3.9x10-6 5.3x10-3 2.2x10-6 2.0x10-5 0 0 1,1,1-Trichloroethane 1.0 1,900 1,900 None 1.2x10-6 1.6x10-3 1.2x10-6 8.5x10-7 0 0 (TCA) Hazard Index - - - - - - 3.7x10-6 2.9x10-5 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-31.-Risk Assessments from Exposure to Hazardous Chemicals from Modular High Temperature Gas-Cooled Reactor Operation at Savannah River Site Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 4.0x10-6 6.0x10-5 6.1x10-8 2.2x10-8 0 0 Ammonia 0.01 27 17 None 1.1x10-6 1.6x10-5 1.1x10-4 6.1x10-7 0 0 Ethanol 46.1 1,900 1,900 None 3.3x10-5 4.8x10-4 7.3x10-7 2.5x10-7 0 0 Methanol 1.75 260 260 None 3.3x10-5 4.8x10-4 1.9x10-5 1.8x10-6 0 0 1,1,1-Trichloroethane 1.0 1,900 1,900 None 1.4x10-6 2.1x10-5 1.4x10-6 1.1x10-8 0 0 (TCA) Trichlorotrifluoroethane 105.0 7,600 7,670 None 1.5x10-4 2.3x10-3 1.4x10-6 3.0x10-7 0 0 (Freon 113) Hazard Index - - - - - - 1.3x10-4 3.0x10-6 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-32.-Risk Assessments from Exposure to Hazardous Chemicals from Advanced Light Water Reactor at Savannah River Site Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetone 10.5 1,800 1,800 None 3.7x10-5 0.05 3.5x10-6 2.8x10-5 0 0 Acetylene 66.28 2,708 2,708 None 1.1x10-5 0.01 1.6x10-7 5.3x10-6 0 0 Ammonia 0.01 27 17 None 1.9x10-5 0.03 1.9x10-3 9.7x10-4 0 0 Ethanol 46.1 1,900 1,900 None 3.9x10-6 5.3x10-3 8.5x10-8 2.8x10-6 0 0 Methane (simple None None None None 1.1x10-5 0.01 0 0 0 0 asphyxiant) Methanol 1.75 260 260 None 3.9x10-6 5.3x10-3 2.2x10-6 2.0x10-5 0 0 Nitric acid 0.123 5 5 None 2.6x10-4 0.35 2.1x10-3 0.07 0 0 1,1,1-Trichloroethane 1.0 1,900 1,900 None 8.5x10-5 0.11 8.5x10-5 6.0x10-5 0 0 (TCA) Hazard Index - - - - - - 4.1x10-3 0.071 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-33.-Risk Assessments from Exposure to Hazardous Chemicals from Accelerator Production of Tritium Operation at Savannah River Site Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 1.1x10-5 0.01 1.6x10-7 5.3x10-6 0 0 Ethanol 46.1 1,900 1,900 None 3.9x10-6 5.3x10-3 8.5x10-8 2.8x10-6 0 0 Methane (simple None None None None 1.1x10-5 0.01 0 0 0 0 asphyxiant) Methanol 1.75 260 260 None 3.9x10-6 5.3x10-3 2.2x10-6 2.0x10-5 0 0 Hazard Index - - - - - - 2.5x10-6 2.8x10-5 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-34.-Risk Assessments from Exposure to Hazardous Chemicals from Tritium Recycling Upgrade Operation at Savannah River Site Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acetylene 66.28 2,708 2,708 None 1.1x10-5 1.4x10-2 1.6x10-7 5.3x10-6 0 0 Ethanol 46.10 1,900 1,880 None 3.9x10-6 5.3x10-3 8.5x10-8 2.8x10-6 0 0 Methane (simple None None None None 1.1x10-5 0.01 0 0 0 0 asphyxiant) Methanol 1.75 260 200 None 3.9x10-6 5.3x10-3 2.2x10-6 2.0x10-5 0 0 Hazard Index - - - - - - 2.5x10-6 2.8x10-5 - - Total Cancer Risk - - - - - - - - 0 0 Table E.3.4-35.-Risk Assessments from Exposure to Hazardous Chemicals from Tritium Recycling Phaseout Function at Savannah River Site [Page 1 of 2] Chemical Regulated Exposure Limits/Risk Factors Emissions Inventory Hazard Quotient Cancer Risk - Reference Permissible Threshold Slope Boundary Worker Boundary Worker Boundary Worker Concentration Exposure Limit Factor MEI 100 Meters MEIa, 100 Meters MEIa, 100 Meters (mg/m3) Limit Value (mg-kg/day) Annual 8 Hours Annual 8 Hours Annual 8 Hours (mg/m3) (mg/m3) (mg/m3) (mg/m3) Acrolein 0.00002 0.25 0.23 None 1.6x10-6 2.1x10-3 7.8x10-2 0.01 0 0 Acrylonitrile 0.002 4.42 4.3 0.24 1.6x10-6 2.1x10-3 7.8x10-4 4.8x10-4 1.1x10-7 2.0x10-5 Antimony 0.00001 0.50 0.5 None 9.2x10-7 1.2x10-3 6.6x10-2 2.5x10-3 0 0 Benzene 0.783 3.25 32.0 0.029 2.3x10-3 3.1 2.9x10-3 0.95 1.9x10-5 3.5x10-3 Cadmium 0.0005 0.2 0.05 None 9.2x10-7 1.3x10-3 1.8x10-3 6.3x10-3 0 0 Cadmium oxide 0.0025 0.1 0.05 None 1.1x10-6 1.4x10-3 4.2x10-4 0.01 0 0 Chlorine 0.039 1.5 1.5 None 1.2x10-4 0.16 3.1x10-3 0.11 0 0 2,4-Dinitrotoluene 0.0069 1.5 0.15 None 1.8x10-5 0.03 2.7x10-3 0.02 0 0 Dioctyl phthalate 0.07 5.0 5.0 0.014 2.1x10-6 2.9x10-3 3.1x10-5 5.8x10-4 8.6x10-9 1.6x10-6 Ethyl benzene 1.0 435.0 434.0 None 2.1x10-5 0.03 2.1x10-5 6.7x10-5 0 0 Ethylene glycol 0.2 740.0 18.0 None 8.1x10-6 0.01 4.1x10-5 1.5x10-5 0 0 Formic acid 6.994 9.0 9.4 None 2.9x10-5 0.04 4.2x10-6 4.4x10-3 0 0 Hexane 0.2 1,800.0 176.0 None 4.1x10-6 5.5x10-3 2.0x10-5 3.1x10-6 0 0 Hydrogen chloride 0.007 7.0 7.0 None 2.9x10-3 3.9 0.41 0.55 0 0 Hydrogen sulfide 0.0009 14.0 14.0 None 8.3x10-5 0.11 0.09 8.1x10-3 0 0 Manganese 0.005 1.0 2.0 None 6.3x10-6 8.5x10-3 1.3x10-3 8.5x10-3 0 0 Mercury 0.0003 0.05 0.05 None 3.2x10-6 4.3x10-3 0.01 0.09 0 0 Methanol 1.75 260.0 200.0 None 6.2x10-6 8.5x10-3 3.6x10-6 3.3x10-5 0 0 Methylene chloride 3.0 1,765.0 177.0 0.0075 2.2x10-5 0.030 7.3x10-6 1.7x10-5 4.7x10-8 8.6x10-6 Methyl ethyl ketone 1.0 590.0 590.0 0 8.1x10-5 0.11 8.1x10-5 1.8x10-4 0 0 (MEK) Methyl tert-butyl ether 0.2 20.0 - 0 3.1x10-5 0.04 1.6x10-4 2.1x10-3 0 0 Methyl isobutyl ketone 0.05 205.0 205.00 0 4.2x10-5 0.06 8.4x10-4 2.8x10-4 0 0 (MIK) Nickel 0.02 1.0 1.0 0.84 4.3x10-6 5.8x10-3 2.2x10-4 5.8x10-3 1.0x10-6 1.9x10-4 Nickel oxide 0.0012 0.1 0.1 0 9.2x10-7 1.2x10-3 7.7x10-4 0.01 0 0 Nitric acid 0.123 5.0 5.0 0 4.8x10-5 0.06 3.9x10-4 0.01 0 0 Sodium hydroxide 0.0489 2.0 2.0 None 6.3x10-6 8.6x10-3 1.3x10-4 4.3x10-3 0 0 Sulfuric acid 0.0245 1.0 1.0 None 9.5x10-7 1.3x10-3 3.9x10-5 1.3x10-3 0 0 Tetrachloroethylene 0.03497 170.0 339.0 None 5.3x10-4 0.72 0.02 4.2x10-3 0 0 (PERC) 1,1,1-Trichloroethane 1.0 1,900.0 1,900.0 None 2.2x10-5 0.03 2.2x10-5 1.6x10-5 0 0 (111-TCA) Trichloroethylene 0.046 270.0 270.0 0.006 1.8x10-4 0.24 3.9x10-3 9.0x10-4 3.1x10-7 5.6x10-5 (TCE) Trichloromethane 0.03497 240.0 9.78 0.081 5.2x10-4 0.70 0.02 2.9x10-3 1.2x10-5 2.2x10-3 (Chloroform) 1,1,2-Trichloro- 105.0 7,600.0 7,670.0 None 1.6x10-5 2.1x10-3 1.5x10-7 2.8x10-7 0 0 1,2,2-Trifluoroethane (Freon 113) Toluene 0.4 375.0 377.0 None 2.8x10-5 0.04 7.1x10-5 1.0x10-4 0 0 Xylene 7.0 435.0 435.0 None 3.8x10-4 0.51 5.4x10-5 1.2x10-3 0 0 Hazard Index - - - - - - 0.7 1.8 - - Total Cancer Risk - - - - - - - - 3.2x10-5 5.9x10-3 Table E.3.4-36.-Risk Assessments from Exposure to Hazardous Chemicals at Savannah River Site-Summary Hazard Index and Total Cancer Risk Technology Hazard Index Total Cancer Risk - Boundary Worker 100 Boundary Worker 100 MEI, Meters MEI, Meters Annual 8 Hoursb Annual 8 Hoursd, No Action 0.70 1.8 3.2x10-5 5.9x10-3 Tritium Recycling Upgrade 2.5x10-6 2.8x10-5 0 0 Heavy Water Reactor 3.7x10-6 2.9x10-5 0 0 Heavy Water Reactor and Tritium Recycling Upgrade 6.1x10-6 5.8x10-5 0 0 Heavy Water Reactor, Tritium Recycling Upgrade and No Action 0.70 1.8 3.2x10-5 5.9x10-3 Modular High Temperature Gas-Cooled Reactor 1.3x10-4 3.0x10-6 0 0 Modular High Temperature Gas-Cooled Reactor and Tritium Recycling Upgrade 1.4x10-4 3.1x10-5 0 0 Modular High Temperature Gas-Cooled Reactor, Tritium Recycling Upgrade, 0.70 1.8 3.2x10-5 5.9x10-3 and No Action Advanced Light Water Reactor 4.1x10-3 0.071 0 0 Advanced Light Water Reactor and Tritium Recycling Upgrade 4.1x10-3 0.071 0 0 Advanced Light Water Reactor, Tritium Recycling Upgrade, and No Action 0.70 1.9 0 0 Accelerated Production of Tritium 2.5x10-6 2.8x10-5 0 0 Accelerated Production of Tritium and Tritium Recycling Upgrade 4.9x10-6 5.7x10-5 0 0 Accelerated Production of Tritium, Tritium Recycling Upgrade, and No Action 0.70 1.8 3.2x10-5 5.9x10-3 Tritium Recycling Only 0.70 1.8 3.2x10-5 5.9x10-3 E.4 Health Effects Studies: Epidemiology Various epidemiologic studies have been conducted in the past at some of the sites evaluated in this PEIS because of the concern with the health effects of nuclear research, and the manufacture and testing of nuclear weapons. These studies are summarized below. With a few exceptions most epidemiologic studies of the populations living near the sites have been descriptive in nature and are what epidemiologists refer to as ecologic or correlational studies. Occupational epidemiologic studies, studies of workers, have been mostly analytical. The various epidemiologic studies, along with their assumptions and limitations, are described in section E.4.1. E.4.1 Background Nuclear weapons research and manufacture and consequent exposure began in the late 1930s and early 1940s. Very little knowledge was available at the time on the potential health effects from radiation exposure. As with many other chemicals, exposure levels to nuclear elements have changed over time with higher levels occurring in the early days of research and manufacture. Epidemiologic studies have been conducted to identify adverse health effects associated with nuclear use. Different epidemiologic methods have been used to assess these health effects; however, the epidemiologic methods for assessing occupational exposures have also evolved over time. For example, the use of multivariate methods for data analysis is a more recent phenomenon. The multivariate methods are more powerful in identifying causality than the ecologic and descriptive methods that were used more fre- quently just a few years ago. Thus, the review of the epidemiologic literature and the resulting findings basically reflect this change in study design over time. E.4.1.1 Study Designs Studies that have been conducted to assess the health effects resulting from exposure to nuclear radiation include ecologic studies, cohort studies, and case-control studies, which include the nested case-control design. The unit of observation in Ecological or Correlational Studies is a group of people rather than an individual (Rothman 1986a). Examples of groups may include classes in a school, factory workers, residents of a city, county, or nation. The health effects measured are frequently limited to the incidence (newly diagnosed cases) or mortality rates because these health indicators are readily available for the unit population under consideration. Exposure is defined as residence in the unit of observation and every person in the unit is assumed to be equally exposed whether they have resided in the unit for a very short time period or a very long time period or whether they even resided in the unit at the time of exposure. Exposure is also assumed to be limited to the unit of observation, a fact that may or may not be true for the 5 candidate sites because of the potential downwind emissions considered to be an important factor in population exposure. The unit of observation for the ecologic studies conducted around the PEIS sites is the county(ies) where the facility is located. Rates in these counties are compared to rates in counties farther away from the site. Because measurements are averaged over populations and because the measures of exposure are only proxies, the ecologic studies tend to attenuate any real associations between exposure and risks unless the risks of disease are very large. No control of confounding is possible in these studies, except for general population rates such as income, race, or crowding. Ecologic studies, however, are frequently used as an overall screening method to assess whether a major health problem exists. Ecologic studies are also relatively inexpensive to conduct compared to other epidemiologic studies because they use readily available data (mortality rates, cancer registry rates, census data) and can be done relatively quickly. Most popu- lation-based studies at the PEIS sites have been ecologic (correlational) studies. The cohort (follow-up) method has been most frequently used to date for the analysis of occupational exposures. Cohort studies include the entire available study population, although some populations may be limited by employment or termination dates (RMOE 1989a). A cohort study requires the enumeration and follow-up of a population of indi- viduals who are known to be exposed or not exposed to an agent of interest (although exposure measurements can be very general or very specific). Cohort follow-up studies can be prospective, where the cohort is identified at the beginning of the study and followed over time, or historical, where the cohort of individuals, most frequently workers, is enumerated as of some time in the past and the health of the cohort is assessed over historical time to estimate disease rates. Most cohort follow-up studies conducted at the PEIS sites have been historical cohort studies. Traditionally, overall and causespecific mortality rates have been assessed. More recently, overall and cause-specific cancer incidence (newly diagnosed) rates have also been assessed. Use of incidence rates, however, is limited to their availability. Rates in the exposed cohort are compared to those observed in a nonexposed population such as the national or regional populations (Standard Mortality or Incidence Ratios), or with subgroups of the cohort classified according to exposure type or level (relative risks, risk ratio, rate ratio). The advantage of the historical cohort design is that it allows follow-up of a large number of individuals (thousands) over a long period of time (years), which would not be feasible for prospective cohort studies (individuals could outlive the investi- gator). But the method also has its drawbacks. Past records are not readily available and may be incomplete or destroyed. Cohort studies are also expensive and time consuming since records for all subjects must be abstracted. The case-control study design reduces cost and time by requiring that information be collected only for the cases and a sample of the cohort that generated the cases (the controls), thus mitigating the difficulties of following a large cohort over time. When used properly, case-control studies provide estimates of the relative risks (odds ratio) of cohort studies (RMOE 1989a). The problem with case-control studies is the identification of controls that are representative of the population that generated the cases. Some case-control studies have been used to study populations around some PEIS sites. The nested case-control design avoids the problems of the general case-control study while taking advantage of its cost saving feature because of the limited number of individuals studied. In a nested case-control study, the enumerated cohort is restricted to the analysis of observed cases and a sample of other individuals free of the disease at the time when the disease in the cases occurred. Several nested case-control studies have been conducted among the workers at the PEIS sites (e.g., ORR). These studies provide estimates of the relative risks (odds ratios) from cohort studies, and the results can be generalized to the entire cohort. E.4.1.2 Definitions Standardization: A method used to control the effects of age, gender, or other known differences when comparing two or more different populations. These methods are used to obtain expected numbers of cases. There are two main methods: Direct Method: The disease rates in the study population are used and are multiplied by the number of individuals in the age-, gender-, or other grouping in the standard population to obtain the expected rates of disease for the study population if the study population had the same age-, gender-, or other characteristics of the standard population. Indirect Method: Because of the small number of individuals in a study population the rates can be very unstable. When this happens, the standardization method of choice is the indirect method. The disease rates in the standard population are multiplied by the number of individuals in the age-, gender-, or other specific group in the study population to obtain the expected rate of disease in the study population if the study population had the same disease rates as the standard population given the study population's age, gender, or other distribution characteristics. Standardized Mortality (Incidence) Ratio: The standard mortality ratio is the ratio of the number of events observed in the study group or population to the number that would be expected if the study group or population had the same rates as the standard (reference) population. These rates have been standardized to the same standard population distribution. Data from the United States population or state-specific populations are most frequently referenced when this rate is used. Standard mortality ratios are frequently used with ecologic and historical cohort follow-up studies. Lack of comparisons between standard mortality ratios from different populations and the healthy worker effect are two distinct disadvantages in using this rate. Standardized Rate Ratio: A rate ratio in which the numerator and the denominator have been standardized to the same (standard) population distribution. Relative Risk, Risk Ratio, Rate Ratio: The ratio of the risk of disease or death among the exposed to the risk among the unexposed. Also, the cumulative incidence rate in the exposed to the cumulative incidence rate in the unexposed. Odd Ratio, also Cross-Product Ratio, and Relative Odds: The odds ratio is the ratio of the odds of getting the disease if exposed, to the odds of getting the disease if not exposed. Under certain conditions the Odd Ratio approximates the Rate Ratio. Healthy Worker Effect: A phenomenon observed in studies of occupational diseases. Workers usually exhibit lower overall death or disease rates compared to the general population, due to the fact that the severely ill and disabled are excluded from employment. Rates from the general population may be inappropriate for comparison if this effect is not taken into consideration. E.4.2 Idaho National Engineering Laboratory Surrounding Communities. A population-based mortality study, carried out by the National Cancer Institute (NIH 1990a; AMA 1991a), examined cancer mortality within a 50-mile radius around several nuclear facilities including INEL. The Idaho study counties included Bingham, Butte, and Jefferson Counties. No excess cancer mortality was observed in the population living in counties surrounding INEL when compared to the United States population, or to the population in the three control counties (Fremont, Cassia, Power, Madison MT, Broadwater MT, Custer, Twin Falls, Lemhi, and Oneida) for each study county, or when time trends were assessed. The study is limited by its correlational approach and the large size of the geographic areas (counties) used for surrogate exposure mea- surements. The Idaho Department of Health and Welfare (Epidemiology and Health Statistics group) completed a different community-based epidemiologic cancer morbidity (1978-1987) and mortality (1950-1989) study of two additional counties (Minidoka and Clark) also within a 50-mile radius near INEL. Clark County lies northeast, and Minidoka County lies southwest of INEL. Clark County population was 880 in 1988, while the 1988 population for Minidoka County was 20,100 (IN DHW 1991a). Cancer death and incidence rates of the Minidoka and Clark populations were compared to those of the entire Idaho population. No differences in age-adjusted mortality rates were observed in either county. The overall annualized cancer death rate for Minidoka County was 314/100,000 compared to 311.3/100,000 for the state (IN DHW 1991b). The cancer death rate for Clark County was 517.3/100,000 but because of the very small number of people living in Clark County this death rate may be very unstable. An excess risk was reported, however, for cancer-specific incidence rates. Clark County had an excess of newly diagnosed radiogenic and nonradiogenic cancers, but of these only breast (8 observed, 3.6expected), which is radiogenic, and lip cancer in males (3 observed, 0.3 expected), which is not radiogenic, were statistically significant. The number of cases, however, was very small and the confidence intervals were wide. Minidoka County also had an excess in radiogenic and nonradiogenic incident cancers; however, only lip (23 observed, 8 expected) and corpus uteri (40 observed, 24.2 expected), both nonradiogenic cancers, reached statistical significance. Since exposure and dose data from INEL were not used for these studies, the excess in cancers are just as likely to be associated with competing causes such as farming and/or smoking and may not be associated with the INEL site specifically. Exposure-related excesses may also be masked due to the imprecise comparisons made, the use of the correlational approach as a surrogate for exposure measurement, the large size geographic areas considered, and the small sample size. Workers. Although no occupational epidemiologic studies have been conducted at INEL to date, according to the the National Institute for Occupational Safety and Health, one is currently underway, but no results are expected before 1997. The Centers for Disease Control and Prevention, as a result of the Memorandum of Understanding with the Department of Health and Human Services of 1991, is currently conducting a dose reconstruction study in anticipation of future epidemiologic community studies. E.4.3 Nevada Test Site Surrounding Communities. Because of the potential carcinogenic threat of radioactive fallout on the general population resulting from above-ground nuclear weapons testing, epidemiologic studies have attempted to assess the cancer incidence and mortality among individuals living in the areas of potential fallout, including areas in Utah. A series of ecological studies showed contradictory results in potentially exposed children. Lyon examined the cancer mortality (leukemia and other) of children living in high and low fallout areas between 1944 and 1975 (NEJM 1979a). Only children born between 1951 and 1958 (period of above-ground testing) were considered as highly exposed if they lived in one of the high-exposure areas at the time of death (Utah was exposed to 26 nuclear tests from the Nevada Test Site (NTS) between 1951 and 1958). Cancer cause was obtained from the death certificates and were separated into leukemias and others. Results show that although Utah children experience an overall cancer mortality rate lower than U.S. children as a whole, children living in high-fallout areas had more leukemia deaths than children living in low fallout areas. Other cancers occurred at the same rate in both high and low fallout areas. The high-risk fallout area identified by Lyon et al. was questioned by Beck and Krey (Science 1983a) who reconstructed the total Cs137 inventory across the state of Utah (area near NTS). Their measurements showed that sections of northern Utah also had high levels of fallout and questioned the results of any comparisons made between residents of the North and those of the South. Land (Science 1984a) also re-examined cancer mortality reported by Lyon (NEJM 1979a) using the National Center for Health Statistics data which covered the period 1950 through 1978. The analyses examined a reported association between childhood leukemia and exposure to radioactive fallout from above-ground nuclear weapons tests in Nevada. The exposure group consisted of 17rural counties in Utah which were downwind from the tests and the control group consisted of the remaining 12 northern counties of Idaho and no- or low-fallout regions outside Utah (Eastern Oregon and the State of Iowa). Although an excess of leukemia was reported for high exposure levels compared to low exposure levels, the data did not support a regional difference because leukemia rates were also significantly elevated in Oregon, an area of no fallout. Another study that assessed the development of cancer among individuals potentially exposed to radioactive fallout has been reported by Rallison (HP1990a). This study examined the thyroid cancer risk in a cohort of children born between 1947 and 1954 in two counties near the nuclear test sites, one in Utah and one in Nevada. A comparison group of Arizona children was also evaluated. The children (11 to 18 years of age) were examined between 1965 and 1968 for thyroid abnormalities and were reexamined in 1985 and 1986. Like the ecological studies, however, exposure was based on geopolitical boundaries. Children living in the nuclear testing area had a higher rate of thyroid cancer but the differences were not statistically significant. Some examined children currently living in the high-fallout area at the time of examination were later found to have been living outside the high-fallout area at the time of exposure. A more definitive study (AMA 1984a) examined the cancer incidence in an area of radioactive fallout downwind from NTS. This study consisted of two parts: an analysis of all Utah Mormons and a personal interview of individuals residing in a high-fallout area. Cancer incidence data for all Utah Mormons was available and two 9-year periods (1958 to 1966 and 1972 to 1980) were selected for the study. The average annual age-adjusted cancer incidence rate was reported as 228/100,000 for all Utah Mormons between 1967 and 1975. The cancers statistically significant at the 1 percent probability level for all Utah Mormons during both time periods were cancers of the thyroid (standard incident ratio (SIR)) (SIR=4.3-8.2), bone (SIR=10.0-12.5), and leukemia (SIR= 5.28-3.5). The ratio of radiosensitive cancer to all other cancers was 53.3 percent higher in the early period (1958 to 1966) and 300percent higher in the later period (1972 to 1980). The lower range of whole-body radiation associated with fallout symptoms in this study is about 50rads. Sixty percent of the 4,125 Mormons who were listed in the telephone directory in 1951 and in 1962 were located and interviewed in 1981. Information on personal habits, lifestyle, cancer, and reproduction was obtained for each family member. This study is one of the few where the investigators were able to interview the study subjects, where the reported cancers were verified, and where a comparable and appropriate comparison group was available. The cancer incidence was greater in this group compared to all Utah Mormons, but most cancers were reported in the period between 1958 and 1968 (stomach, leukemia, brain, and melanoma) and only a few (breast and lymphoma) were statistically significant in the later time period (1972 to 1980). This part of the study would have been more powerful if the investigators had interviewed a comparable group of controls. Cancers of all sites were significant at the 1 percent probability level for all groups studied and for all time periods under consideration. Machado (AJE 1987a) examined the cancer mortality experience of a three-county region in southwestern Utah using a correlational cross sectional mortality study and compared the cancer mortality there with that from the rest of Utah. This study confirmed the excess leukemia reported by Johnson (AMA 1984a) but could not confirm the excess cancers at other sites. Because the studies have different designs they cannot easily be directly compared. Archer (AEH 1987a) measured soil, milk, and bone strontium-90 levels to identify states with high, intermediate, and low fallout contamination. He then correlated the deaths from radiogenic and nonradiogenic leukemias with the time periods of above-ground nuclear testing both in the United States and Asia. The results show that leukemia deaths in children were higher in states with high exposure and lower in states with less exposure. He showed that leukemia deaths in children peaked approximately 5.5 years following nuclear testing peaks. The last leukemia peak in the United States occurred in 1968 to 1969, 5.5 years after the last year of a 3-year period of intensive testing in Asia. The increases were seen in the radiogenic leukemias (myeloid and acute leukemias), and not with "all other leukemias." Workers. An excess number of leukemia cases was reported (9 cases, 3.5 expected) among the 3,224men who participated in military maneuvers in 1957 at the time of the nuclear test explosion "Smoky" (AMA 1980a; AMA 1983a). The participants were located and queried on their health status, diseases, or hospitalizations as of December 1981. A next of kin was queried about cause of death for those participants who were deceased. Exposure information was available from film badges and ranged from 40 to 1,834 mrem. In the later report (AMA 1983a), the number of incident cases of leukemia had increased to 10 with 4 expected. No excess in "total cancers" was observed, however. In addition, four cases of polycythemia vera were reported where 0.2was expected (AMA1984a). The excess in leukemia cancer incidence and mortality appear to be limited to the soldiers who participated in "Smoky." The excess was not observed in a National Research Council mortality study of soldiers exposed to five series of tests at two sites: Nevada Test Site (PLUMBBOB) and the Pacific Proving Ground (National 1985a). The NRC reported that the number of leukemia cases in "Smoky" was greater but the increase was considered nonsignificant when analyzed with the data from the other four tests. Re-analysis of the National Research Council report on the Mortality of Nuclear Weapons Test Participants released on June 4, 1985, showed a 62-percent excess cancer among soldiers involved in the nuclear test explosion PLUMBBOB ("Smoky" was part of PLUMBBOB) who were exposed to 300mrem or more (AJE 1987a). The excess was observed for leukemia and aleukemia as well as the other cancer group. The original analyses acknowledged the healthy worker bias when the expected mortality values were obtained from the general population but the method of analysis did not control for the problem in the original study. The reanalysis avoided the problem by considering a dose response analysis. Two exposure groups were considered: 300mrem and <300 mrem. The excess was observed in the group exposed to more than 300 mrem. The reanalysis also reports an elevated risk of death from respiratory cancers for "Smoky" participants (Cancer Mortality Ratio =2.28, p <0.01) and an excess risk of "other" (Ratio=1.75, p <0.01) and "all cancers" (Ratio=1.49, p <0.01) for the other participants of "PLUMBBOB." E.4.4 Oak Ridge Reservation More epidemiologic studies have been carried out to assess the health effects of the population living near or working at the ORR facilities in Tennessee than at any other site reviewed for the PEIS. Surrounding Communities. The population-based National Cancer Institute mortality survey for selected nuclear facilities (NIH 1990a; AMA 1991a) examined the cancer mortality within a 50-mile radius around several nuclear facilities including Anderson and Roane counties. No excess cancer mortality was observed in the population living in the exposed counties when compared to the United States white male population, nor when compared to the population of the control counties (Blount, Bradley, Coffee, Jefferson, Hamblen, TN, and Henderson, NC), nor when time trends were assessed. The Tennessee Cancer Reporting System (OR DHE 1992a) of the Tennessee Department of Health and Environment also evaluated the cancer incidence and cancer mortality rates for Anderson and Roane counties and compared them with the average rates of 72 U.S. cities. Slight excesses in cancer incidence and overall mortality rates were observed, none of which were statistically significant. For Anderson County, there was a slight excess in the incidence of myelomas (observed to expected (O/E) ratio is 1.58), and cancers of the lung and bronchus (O/E=1.21), esophagus (O/E=2.19), and cervix uteri (O/E=1.89). For Roane County, a slight excess in the incidence of cancers of the bone and joints (O/E=2.50), lung and bronchus (O/E=1.40), oral cavity and pharynx (O/E=1.59), and stomach and small intestines (O/E=1.27) was reported. Mortality rates were elevated for respiratory cancers (O/E=1.24) and for cancers of the skin (O/E=2.6) in Anderson County, and for cancers of "other lymphatic and hematopoietic tissue" (O/E=1.45), respiratory and intrathoracic organs (O/E=1.32), and for unspecified and "all other cancers" (O/E=1.26) in Roane County. The only consistent excess reported for both cancer incidence and cancer mortality rates occur with cancer of the respiratory and intrathoracic organs, an excess which may be confounded by smoking. The two population-based studies mentioned above are limited by their correlational approach and the large size of the geographic areas (counties) used. Because of a concern for possible contamination of the population by mercury, the Tennessee Department of Health and Environment conducted a pilot study in 1984 (OR DHE 1984a). The study showed no difference in urine or hair mercury levels between individuals with potentially high mercury exposures (residence or activity in contaminated areas based on soil measurements or consumption of fish caught in the contaminated areas) compared to those with little potential exposure. Mercury levels in some soils measured as high as 2,000 parts per million (ppm). Analysis of a few soil samples showed that most of the mercury in the soil, however, was inorganic thereby lowering the probability of bioaccumulation and health effects. Examination of the long-term effects of exposure to mercury and other chemicals continues. Workers: Radiation. Between 1943 and 1985, there were 118,588 male and female individuals of all races who were ever employed in any of the Oak Ridge facilities. These included the Oak Ridge National Laboratory for nuclear research (also called the X-10 facility), Y-12 plant under management of the Tennessee-Eastman Corporation (1943 to 1947) which produced enriched uranium by the electromagnetic separation process, Y-12 under management of Union Carbide (1948 to 1984) which fabricated and certified nuclear weapons parts, and K-25 (Oak Ridge Gaseous Diffusion Plant) which produced enriched uranium through the gaseous process. Analyses at the Oak Ridge facilities have been carried out mostly for white males, and for specific cohorts taking into consideration time-related exposure risks. All mortality rates were compared to that of United States white males unless otherwise specified. Oak Ridge National Laboratory. An excess of leukemia deaths (standard mortality ratio (SMR)) =1.48; 95 percent Confidence Interval (CI)=0.31-4.38) among 8,375 white male Oak Ridge National Laboratory workers employed between January 1, 1943, and December 31, 1972, was reported by Checkoway et al., (1985). The study excluded workers known to have worked at other nuclear facilities. Data from film badges used since 1951 provided an assessment of external penetrating radiation exposure. An elevated SMR was also reported for prostate cancer and Hodgkin's disease. No dose response, however, was observed for prostate cancer and Hodgkin's disease, and the numbers were too small to be statistically meaningful. The excess leukemia deaths reported in this study occurred mostly among maintenance workers and engineers who had worked for more than 10years, suggesting a possible excess attributed to exposures other than radiation. Maintenance workers are exposed to a number of solvents, metals, and welding fumes. No attempt was made in this analysis to control for confounders such as smoking and socioeconomic status. The excess of leukemia deaths reported by Checkoway et al. (BJIM 1985a) was confirmed by Wing (AMA 1991a; AJIM 1993a) when the cohort of white males (n=8,318) was followed through 1984 (SMR=1.63; 95 percent CI=1.08-2.35 for all workers; SMR=2.23; 95 percent CI=1.27-3.62 for workers monitored for internal contamination). An estimated percentage increase for leukemia rose from 6.38 percent for an exposure lag of 0years to 9.15 percent with a 10-year lag. Of the 30leukemia deaths in the cohort, 2 (0.7 expected) occurred in the radioisotope group, 2 (0.5 expected) occurred in the chemical operations group, and 5 (2.3expected) occurred in the engineering research and development group. An increase of 2.68 percent in deaths from "all causes" and 4.94 percent for "all cancers" with every 10 million Sieverts (1 rem) of cumulative dose exposure with a 20-year exposure lag was also reported by Wing (AMA 1991a;AJIM 1993a). Adjustments for confounders such as job category; socioeconomic status; World War II employment; and exposure to beryllium, lead, and mercury had little or no effect on the excess risk observed. Risk estimates between cumulative exposure to low levels of external penetrating ionizing radiation and "all cause" mortality were shown to be several times the risk estimates obtained from the Japanese A-bomb survivor data for low-level radiation. This excess increased with increasing dose regardless of exposure lag time, and was stronger for exposures that occurred 20 years in the past. The excess deaths due to ill-defined causes (SMR=2.34; 95 percent CI=1.76-3.05 for all workers; SMR= 2.89; 95 percent CI=1.93-4.14 for workers monitored for internal contamination) was also statistically significant. These increases remained consistent regardless of whether confounders (active work, pay code, cohort, and continuous or categorical age) were considered. Deaths from lung cancer increased 5 percent for every 10 million Sieverts (1 rem) of exposure (AMA 1991a; AJIM 1993a). The excess in cancer deaths was associated with working in radioisotope production and chemical operations but not with work in physics, engineering, or unknown job categories. Cancer mortality was also associated with exposure to beryllium, lead, and mercury. Control of these chemicals, as well as the job exposure variable in a Poisson regression model, did not significantly affect the initial dose-response relationship observed between external low-level radiation and leukemia risks. Y-12 Plant. Internal radiation exposure by inhalation of uranium compounds was the major concern for workers at the Y-12 Plant (Y-12) under Union Carbide management employed between 1947 and 1969 (n=4,988) and followed to 1973. An elevated but not statistically significant SMR for cancer of the brain (SMR=1.97 with 7 observed and 3.56expected; 95 percent CI=0.79-4.06) was reported in this cohort when compared to U.S. white males (Polednak 1980a). The effect of alpha (internal) and gamma (external) radiation was reported by Checkoway (AJE 1988a). The cohort was restricted to 6,781 white males employed for at least 30 days and who worked between May 4, 1947, and December 12, 1974, at the Y-12 facility. Employees who worked at another nuclear facility and any worker employed prior to 1947 were excluded from the analyses (a major shift from uranium enrichment to fabrication of weapons' parts and research and development for separation of isotopes of other compounds occurred in 1947). Exposure data were available from film badges and urinalyses. Urinalysis data were converted to estimates of lung dose equivalents through the use of metabolic models. Observed mortality rates were compared to those from the U.S. white male population and to that of Tennessee white males. An excess of lung cancer mortality was observed in this cohort (SMR=1.36; 95 percent CI=1.09-1.67 when compared to the U.S. white male rates; SMR=1.18; 95 percent CI=0.95-1.45 when compared to rates from Tennessee white males). A dose-response effect appeared related to the alpha (internal) and gamma (external) radiation with the most pronounced effect observed for gamma radiation among workers exposed to 5 rems of alpha radiation. At zero-year latency assumption, the dose-response was very pronounced for workers with both 5 rems or more of alpha and gamma radiation (Rate Ratio (RR)=4.60; 95 percent CI=0.91-23.4). The effect was weaker when a latency of 10 years was considered (RR=3.05; 95 percent; CI=0.37-24.83), suggesting that the effect may not be related to radiation. However, the estimates were based on a small number of cases (3 deaths at zero-year latency and 1death at 10 ten-year latency) and consequently lack statistical precision. The period of observation would need to be extended for the study to have sufficient power to detect a moderate effect. Checkoway (AJE 1988a) also reported an excess of brain and central nervous cancers in this cohort (SMR=1.80; 95percent CI=0.98-3.02 when compared to rates of U.S. white males; SMR=1.42; 95 percent CI=0.78-2.38 when compared to rates of Tennessee white males) but no dose response relationship was observed. Polednak also reported on the mortality experience of white male workers employed at the Y-12 facility while under Tennessee-Eastman Corporation management between 1943 and 1947 (Polednak 1980a). These workers (n=4,008) were followed through 1973. The cohort excluded workers who continued employment after the shift in 1947 and those who worked for any other ORR facility. Exposure to external penetrating and internal radiation from uranium dusts was of concern for the Tennessee-Eastman Corporation workers. Compared to U.S. white males, the SMRs for white Tennessee-Eastman Corporation males who worked 1 year or longer was elevated for cancer of the testis (1.92), benign neoplasms (1.45), all diseases of the bones and organs of movement (3.22; 4 of the 6 deaths were due to rheumatoid arthritis), and "symptoms, senility, and ill-defined conditions" (1.77). Although the investigator reported that the 95 percent confidence estimates were calculated, none were presented in the tables. An elevated risk for brain cancer was reported among 4,988 white males (Pedant 1980a) employed between 1947 and 1969 (Y12 workers under Union Carbide) but the excess was not statistically significant (SMR=1.97; 95 percent CI=0.79-4.06). A follow-up through 1974 of all 18,869 workers employed at the Y-12 facility while under the TEC management between 1943 and 1947 was reported by Pedant and Frome (JOM 1981a). The workers included those exposed to internal ("alpha"), external ("beta") radiation through the inhalation of uranium dusts, electrical workers who performed maintenance in the exposed areas, and other non-exposed workers. Elevated but not statistically significant SMRs were observed in this cohort for mental, psychoneurotic, personality disorders (SMR=1.36), diseases of the bones and organs of movement (SMR=1.22), and symptoms, senility, and ill-defined conditions (SMR=2.80). No elevated SMRs were observed for cancer of the bones and for leukemia. The crude SMR for lung cancer was 1.09 (95 percent CI=0.97-1.22) but when adjusted for incomplete retrieval of death certificates, the SMR was significant (SMR=1.22; 95 percent CI=1.11-1.36). Cancer of the lung was greater among workers employed for 1 year or more compared to workers employed less than 1 year, and was more pronounced in workers hired at the age of 45 or older (odds ratio=1.51; 95 percent CI=1.01-2.31). Of the workers employed after the age of 44, the SMR for lung cancer was greater for electrical workers (SMR =1.55) confirming the excess observed in the earlier analysis, for alpha chemistry workers (SMR=3.02), and for beta process workers (SMR=1.51). Combined Oak Ridge National Laboratory Facilities. Frome reported on the mortality experience of World War II workers employed at three ORR facilities between 1943 and 1947 using poisson regression analyses to control for potential confounders (such as facility TEC, K-25, X-10, or multiple), socioeconomic status using pay code (hourly, monthly), length of employment, period of follow-up, and birth earlier 1990a). The cohort included only white males who worked at ORR at least 30 days between the start of the operation and the end of the study (December 31, 1947), and at no other time after that date. The regression analyses confirmed the significantly elevated standard mortality ratios observed for deaths from tuberculosis (SMR=1.37), from mental, psychoneurotic, and personality disorders (SMR=1.60), and symptoms of senility, ill-defined conditions (SMR=3.05) reported by Polednak (1980a). No deaths were associated with exposure to radiation, however. The regression analysis showed a significant association with socioeconomic status as measured by job skill. Carpenter investigated the earlier reports of an association between the excess number of brain cancers and employment at the Oak Ridge National Laboratory and the Y-12 Plant in a case-control study among workers employed between 1943 and 1977 (AJIM 1988a). Cases consisted of 72 white males and 17 white females. Four controls were selected for each case and matched on age, sex, cohort, year of birth, and year of hire. No statistically significant association between the use of 26 chemicals evaluated and an increased risk of brain cancer was observed. The chemicals evaluated included those encountered in welding fumes, beryllium, mercury, 4,4'-methylene bis 2-chloroaniline or MOCA, cutting oils, thorium, methylene chloride, and other solvents. However, the excess brain cancer was observed among individuals employed for more than 20 years (oddsratio =7.0; 95 percent CI=1.2-41.1; cases=9). Analyses of these cases and controls with respect to internal and external radiation exposures revealed no association between radiation exposures and development of brain cancer (JOM 1987a). However, analysis of 82 cases with complete medical records revealed an excess risk of brain cancer with a history of epilepsy (odds ratio=5.7; 95 percent CI=1.0-32.1) as recorded in the medical records for pre-employment and health status follow-up (AJPH 1987a). Brain cancer of glial origin (4 cases and 0 controls) was strongly associated with a history of epilepsy. Combined Nuclear Sites. ORR workers were also included in a combined analysis of Hanford Site, ORR, and Rocky Flats Plant (now known as Rocky Flats Environmental Technology Site) workers (RR 1989a). White male workers employed at one of the three facilities for at least 6 months and who were monitored for external radiation were included in the study. Analyses were based on the same vital status and cause of death information previously reported. Deaths were considered through 1981 for Hanford, through 1977 for ORR, and through 1979 for Rocky Flats Plant. Hanford contributed 23,704 workers, ORR contributed 6,332 workers, and Rocky Flats Plant contributed 5,897 workers. External radiation doses were obtained from the dosimeters worn by these workers. A check for possible overlap of the Hanford and Rocky Flats Plant cohorts identified 78workers who qualified in both study populations. These workers were included in the cohort of the facility where they first met eligibility and the dose accumulated at the second facility was not included. These analyses provide no evidence of a correlation between radiation exposure and mortality from all cancer or from leukemia and no evidence of chronic low dose exposure with any form of cancer mortality at ORR. Multiple myeloma was the only cancer found to exhibit a statistically significant correlation with radiation exposure in this multi-site analysis. However, all cases were contributed by the Hanford cohort (i.e., none was from ORR). Workers: Chemicals and Metals. Only a few studies have examined the mortality of ORR workers exposed to other chemicals or metals. Polednak reported elevated mortality rates for white male workers exposed to phosgene while employed at Y-12 between 1943 and 1947 and followed through 1974 (ER 1980a). The SMR for symptoms, senility, and ill-defined conditions was 4.02 (95 percent CI=1.61-8.28) compared to the United States white males. The SMR, however, was also elevated among workers not exposed to phosgene (SMR=3.23; 95 percent CI=2.62-3.97). Although not statistically significant, the analysis shows an elevated SMR for all infective and parasitic disease (SMR=1.22; 95 percent CI=0.4-2.84) among workers with low to moderate phosgene exposure, and which was not observed among workers not exposed to phosgene (SMR=0.80; 95 percent CI=0.62-1.04). The excess was due mostly to tuberculosis among exposed workers who had no prior history of tuberculosis. Among workers highly exposed to phosgene, an elevated but not statistically significant SMR for external causes of death (accidents) was also reported (SMR=1.62; 95 percent CI=0.52-3.77). This designation (external causes of death) includes acute chemical poisoning and at least one of the four deaths is known to have resulted from phosgene poisoning. The investigator expresses concern about one death attributed to paranoid schizophrenia since dementia precox (schizophrenia) associated with phosgene exposure was previously reported in the literature. With only 106 heavily exposed workers, the numbers are too small to identify any trends in mortality. The results of this study also highlight the problem when analyses focus only on mortality and not morbidity. Cragle reported on the mortality experience of 2,133 Y-12 workers exposed to elemental mercury vapors used in the lithium isotope separation process (JOM1984a). Exposure for these workers was documented through urinalysis measurements. Another 270workers were employed in the same departments as those who were monitored and, therefore, were probably exposed. Finally, 3,530 workers were never exposed. The mortality of all groups was compared to the mortality experience of U.S. white males. No statistically significant excess mortality was observed for workers exposed to mercury. The excess mortality observed for cancer of the brain and the lung was observed for all three groups and cannot be attributed to mercury exposure. Why the exposed and nonexposed were not compared using Poisson regression model or a contingency table analysis is not known but may have shed some light on mercury-specific effects. The investigators do point out, however, that mercury may exert its effects more on morbidity than on mortality. Cragle reported on the mortality experience of the workers at the Oak Ridge Gaseous Diffusion Plant (IARC 1984a). Within this group, 800 white males were exposed to nickel powder and 7,500 workers were never exposed to nickel or the "barrier" material. Deaths in the exposed and nonexposed were compared to the mortality experience of U.S. white males. Statistically significant SMRs were observed for symptoms, senility, and ill-defined causes for both exposed (SMR=5.93; 95 percent CI=2.96-10.61) and nonexposed (SMR=3.25; 95percent CI=2.54-4.10) workers and consequently cannot be attributed to nickel exposures. Elevated but not statistically significant SMRs were observed for cancers of the buccal cavity, liver, pancreas, and skin for the exposed workers and but not for the unexposed workers. When direct adjustments were made, all SMRs were lower than expected but the SMRs were higher in the exposed group for cancers of the buccal cavity, liver, and pancreas. E.4.5 Pantex Plant Surrounding Communities. Incidence data from the residents of the southwestern quadrant of Carson County (zip code area 79068) nearest Pantex, which was collected between 1980 and 1990, were compared with age-, gender-, and race-specific cancer incidence in five Public Health Regions of Texas that had complete cancer reporting since 1976. This study was carried out by the Texas Department of Health (PX DOH 1992a; PX DOH 1992). The Standardized Incidence Rate for the overall cancer incidence in the Carson County area of interest was 0.84 for males (72 cancers were reported; 85.75 were expected) and 0.88 for females (63 cancers were reported; 71.96 were expected). The only cancers with elevated but statistically nonsignificant Standardized Incidence Rates were cancer of the brain (SIR=1.82; 95 percent CI=0.22-6.57 in females with 2 cases observed and 1.1 expected) and leukemia (SIR=1.32; 95 percent CI=0.36-3.37 in males with 4 cases observed and 3.04 expected; SIR=2.11; 95percent CI=0.69-4.92 in females with 5 cases observed and 2.37 expected). The wide confidence intervals indicating a decrease in precision can be explained by the small number of cases. Mortality data were obtained from the Bureau of Vital Statistics and consisted of all who died of cancer between 1980 and 1991 in the zip code of interest. The comparison group for the mortality study consisted of age-, gender-, and race-specific rates for Texas during the years 1981-1989. The results for the mortality study were similar to those observed for the incidence study, with no significant increases in cancer mortality identified. Workers. Acquavella compared the mortality experience of Pantex workers employed between 1951 and 1978 with that of U.S. white males (HP 1985a). Exposure information on workers was available from film badges, but only since 1963. Only data for white males (3,564) were reported because of the small number of females (892) and nonwhites (154). The results show a possible healthy worker effect. The Standard Mortality Ratio (SMR) for "all cause" mortality was significantly less than expected (SMR=0.72; 95 percent CI=0.64-0.85) with 269observed deaths and 373.71 expected. The same was true for all the cause-specific deaths except brain cancer (SMR=1.36; 95 percent CI=0.37-3.47) and leukemia (SMR=1.28; 95 percent CI=0.35-3.27). The numbers observed (brain=4; leukemia=4) and the numbers expected (brain=2.95; leukemia=3.13) were so small, however, that the excess could be attributed to chance alone. This study has limitations that restrict the conclusions that can be drawn. First, the comparison group (U.S. white males) does not take into consideration the healthy worker effect. A more appropriate comparison group would include nonexposed workers, or workers exposed at various radiation levels. Secondly, the cohort was followed on the average for only 14.6 years. Except for the leukemias, most cancers generally have a latency period (timebetween cancer initiation and diagnosis) of 20to 40 years. The period of follow-up is inadequate to identify an excess in cancer mortality except for leukemias, and the sample size is inadequate to determine any but the largest differences. E.4.6 Savannah River Site Surrounding Communities. The Savannah River Site, established in 1953 in Aiken, SC, produces plutonium and tritium and other nuclear materials. There are reports that millions of curies of tritium have been released over the years both in plant exhaust plumes and in ground and surface water streams (ED 1982a). Only two epidemiologic studies have been done at the Savannah River Site. They are summarized below. In 1979, Sauer (as reported by Johnson, 1982) reported an excess of leukemia and lung cancer deaths among individuals living within a 50-mile radius of SRS compared to whites living between 50and 99 miles away. Other excesses in mortality were observed but were statistically nonsignificant. Comparison groups in Sauer's study consisted of U.S. death rates and rates for counties more than 50miles from SRS. More current correlation analyses of cancer incidence and deaths reveal no excess leukemia and no excess of other cancers. The most comprehensive of the recent studies was the National Cancer Institute Study (NIH 1990a; AMA 1991a) that examined cancer mortality around nuclear facilities. Exposure was based solely on geopolitical boundaries. Barnwell County was considered the exposed county and Chester, Georgetown, and Sumter counties were considered control counties. The analyses compared the mortality experience in exposed counties with the mortality experience of nonexposed counties and with the mortality experience of U.S. white males. In addition, the mortality experience in the exposed counties occurring before the facility was built was compared with the mortality experience in the same counties after the facility was built. No excess cancer mortality was observed in the population surrounding any facility when compared to the control counties, the U.S. rates, or when the time trends were assessed. Other community studies are being planned. A tumor registry has been in operation for a few years in South Carolina and Georgia, which covers the area around the site. Emory University investigators are currently analyzing data from the Georgia side of the plant while investigators at the University of South Carolina are analyzing data on the South Carolina side. Workers. Evidence of an excess number of leukemia deaths has been reported in hourly workers at the Savannah River Plant (AJIM 1988a). Retrospective and prospective epidemiologic studies are being planned as part of a survey of plutonium workers at four Department of Energy facilities (LosAlamos National Laboratory, Mound Plant, Rocky Flats Plant, and SRS).
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