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Recovery and Transmutation of Iodine-129 in an Accelerator-Driven Transmutation System

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Recovery and Transmutation of Iodine-129 in an Accelerator-Driven Transmutation System WM’01 Conference, February 25-March 1, 2001, Tucson, AZ RECOVERY AND TRANSMUTATION OF IODINE-129 IN AN ACCELERATOR-DRIVEN TRANSMUTATION SYSTEM James C. Bresee U.S. Department of Energy James J. Laidler Argonne National Laboratory, Kimberly W. Thomas Los Alamos National Laboratory ABSTRACT Extensive repository system performance analyses have been carried out in the course of evaluating the feasibility of the proposed Yucca Mountain repository. These studies have shown that the primary contributors to public radiation dose as a consequence of release of radionuclides from the repository are isotopes of iodine, technetium and neptunium (129I, 99Tc, and 237Np). The United States Department of Energy is presently conducting a science-based R&D program, the Advanced Accelerator Applications Program, that addresses the technical issues associated with the transmutation of transuranic elements and long-lived fission products contained in the U.S. inventory of commercial Light Water Reactor (LWR) spent fuel. One of the main issues is the recovery and transmutation of iodine. The two fuels from which the iodine must be extracted are quite different, one (the LWR fuel) being an oxide fuel with zirconium alloy cladding, and the other most likely being a dispersion of metallic transuranics in a metallic zirconium matrix with stainless steel cladding. An aqueous solvent extraction process similar to the PUREX process will probably be used for LWR spent fuel processing, while a pyrochemical process seems to be most appropriate for the transmuter fuel. An overall system recovery target for iodine of 95% has been set as an initial goal of the program. Prior experience with iodine recovery in the PUREX process, both domestic and international, suggests that a satisfactory means can be developed for recovering the iodine contained in the LWR spent fuel. Iodine recovery in the less well-defined pyrochemical process for transmuter fuel appears to be feasible, but experimental verification and validation of the recovery steps is required. It is possible that the form of the recovered iodine from the two fuel types will be different, so it then remains to develop processes for incorporating the iodine in a standard target form, expected to be sodium iodide. Target design will also take into account means for accommodating the products of iodine and sodium transmutation. INTRODUCTION Transmutation by neutron irradiation has been studied for many years throughout the world and is currently under active investigation by the U.S. Department of Energy, within the Advanced Accelerator Applications (AAA) program of the Office of Nuclear WM’01 Conference, February 25-March 1, 2001, Tucson, AZ Energy, Science and Technology. The purpose of this paper is to describe the AAA transmutation work as applied to 129I. CURRENT U.S. HIGH-LEVEL NUCLEAR WASTE MANAGEMENT POLICY The current high-level nuclear waste management program in the United States is based on policies established by the Nuclear Waste Policy Act of 1982, as amended. The Act states that the long-term management of such high-level waste will be through deep geologic disposal. In the case of civilian power reactors, the disposal form is to be untreated spent fuel suitably packaged. The Act also describes a rather intricate process for identification and selection of a disposal site, which through the end of calendar year 1987 produced an initial selection of nine possible sites; this was subsequently reduced to five and finally to three possible sites. In December 1987, the process was modified through an amendment of the 1982 law, resulting in the identification of only one of the candidate sites, Yucca Mountain, Nevada, for study or “characterization”. The Yucca Mountain site is composed of welded and unwelded volcanic ash, called “tuff” to a very great depth. It lies in a desert region on Federal land approximately 170 kilometers northwest of Las Vegas, at the western edge of the Nevada Test Site. One characterization method has been “performance assessment”. The process involves the development of a series of mathematical models based on experimentation and natural analogues which describe the multistage process for movement of radioactivity from the disposal site to the accessible environment. In the case of Yucca Mountain, the accessible environment is defined as a region of marginal farming in Amargosa Valley some twenty to thirty miles southwest of the disposal site where agriculture based on irrigation is currently practiced by several thousand residents. Hydrologic studies have shown that ground water approximately 600 meters under Yucca Mountain flows very slowly to the southwest under Amargosa Valley and reaches the surface in Death Valley where it evaporates, depositing its dissolved minerals and salts in dry lake beds and other deposits. It is an important feature of the Yucca Mountain site that the regional hydrology is “closed”; that is, there is no connection to “open” systems like the Colorado River basin which reaches the ocean. Assuming that Yucca Mountain were selected and licensed for disposal (a process which, if successful, will take about a decade), the multistage process for radioactivity release would include the following steps. First, rain would fall on the mountain, 95% of which would evaporate or run off. The remaining 5% would seep approximately 300 meters through the unsaturated zone until it encounters corrosion-resistant waste packages. Over geologic time (tens of thousands of years) such packages would begin to fail and soluble radionuclides would migrate into the unsaturated flow, eventually reaching the water table some 300 meters below. During that transport, some or most of the radionuclides would be absorbed on the underlying rock. The subsequent saturated flow to Death Valley would also involve absorption as well as dilution from rainfall along its path. Quite complex and integrated models allow a prediction, within bounding limits, of the potential exposure of Amargosa Valley residents to the kind and concentrations of radionuclides in future irrigation waters. WM’01 Conference, February 25-March 1, 2001, Tucson, AZ A series of Total System Performance Assessments (or TSPA’s ) have been carried out for the Yucca Mountain site, in part to evaluate various repository design options but also to determine if the site can meet regulatory requirements. During 2001, the U.S. Department of Energy expects to issue a Site Recommendation Report for Yucca Mountain and an Environmental Impact Statement for the site, containing the latest TSPA information. Although not final at the time of this paper, the preliminary results show a continuation of the same trends that have been reported in previous TSPA’s. That is, from a potential human exposure standpoint, the most important radionuclides during the first 10,000 to 30,000 years of waste storage are 99Tc (213,000 year half- life) and 129I (15.7 million year half-life). Over the subsequent hundreds of thousands of years, 237Np (2.14 million year half-life) is the dominant radionuclide. The final approval of the Yucca Mountain site as a candidate for Nuclear Regulatory Commission licensing and the successful result of the licensing process will depend on whether the predicted performance of a Yucca Mountain repository is acceptable. The next few years will provide answers to that question. TRANSMUTATION In the future and as a possible improvement to current nuclear waste management practices, it is potentially possible to convert, through transmutation reactions, long-lived radionuclides such as radioiodine to short-lived or even stable isotopes. The process involves neutron irradiation, either in a nuclear reactor or in an accelerator-driven subcritical assembly. Long-lived actinides would be eliminated by fission, producing additional fission products for disposal or, if long-lived and mobile, for possible transmutation. Most iodine fission products have short half-lives. After several years, the only remaining iodine in spent nuclear fuel will be stable 127I and very long-lived 129I. The probability of neutron capture by either nucleus (its capture cross section) is a function of neutron energy. Assuming the proper energy is attained, capture in either isotope would produce an unstable nucleus which would decay by beta emission to xenon. Stable 127I would be transformed to 128I, which would beta-decay with a 25-minute half-life to stable 128Xe; 129I would be transformed to 130I, which would beta-decay with a 12-hour half-life to stable 130Xe. Designing a program to collect fission product iodine, transform it to a suitable irradiation target, irradiate it with neutrons of the proper energy, chemically process the targets to recover untransmuted iodine for recycle and stable xenon for release, and repeat the process as needed is a challenging undertaking. CHEMICAL PROCESSING OF SPENT FUEL AND THE RECOVERY OF IODINE FOR TRANSMUTATION Each metric ton of LWR spent fuel, upon discharge from a reactor, contains almost 200 grams of 129I, together with another 45 grams of the stable iodine isotope 127I and about 10 grams of 131I. The 131I isotope poses the greater health concern, should it be released as for example in the case of a reactor accident; but in the case of fuel processing the short half-life (~8 days) of 131I means that this isotope has decayed away before fuel WM’01 Conference, February 25-March 1, 2001, Tucson, AZ processing begins. In the early days of nuclear fuel reprocessing, radioiodine was vented to the atmosphere, and in the rare events when fuel was processed after only very short cooling times, on the order of 15-20 days, the impact to the environment was significant (1). In conventional aqueous solvent extraction reprocessing plants, a small fraction of gaseous iodine is released when the irradiated fuel rods are chopped prior to nitric acid leaching of the fuel. Upon dissolution, much of the iodine present in the fuel is oxidized to the elemental iodine state and appears as a gas; some is present as HI and some as HIO or HIO3.
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