The removal of volatile radionuclides generated during used nuclear fuel reprocessing in the US is almost certain to be necessary for the licensing of a reprocessing facility in the US. Various control technologies have been developed, tested, or used over the past 50 years for control of volatile radionuclide emissions from used fuel reprocessing plants. The US DOE has sponsored, since 2009, an Off-gas Sigma Team to perform research and development focused on the most pressing volatile radionuclide control and immobilization problems. In this paper, we focus on the control requirements and methodologies for85Kr and129I. Numerous candidate technologies have been studied and developed at laboratory and pilot-plant scales in an effort to meet the need for high iodine control efficiency and to advance alternatives to cryogenic separations for krypton control. Several of these show promising results. Iodine decontamination factors as high as 105, iodine loading capacities, and other adsorption parameters including adsorption rates have been demonstrated under some conditions for both silver zeolite (AgZ) and Ag-functionalized aerogel. Sorbents, including an engineered form of AgZ and selected metal organic framework materials (MOFs), have been successfully demonstrated to capture Kr and Xe without the need for separations at cryogenic temperatures.
SUMMARYAs a result of fuel reprocessing, volatile radionuclides will be released from the facility stack if no processes are put in place to remove them. The radionuclides that are of concern in this document are 3 H, 14 C, 85 Kr, and 129 I. The question we attempt to answer is how efficient must this removal process be for each of these radionuclides? To answer this question, we examine the three regulations that may impact the degree to which these radionuclides must be reduced before process gases can be released from the facility. These regulations are 40 CFR 61 (EPA 2010a), 40 CFR 190(EPA 2010b), and 10 CFR 20 (NRC 2012), and they apply to the total radonuclide release and to the dose to a particular organ -the thyroid. Because these doses can be divided amongst all the radionuclides in different ways and even within the four radionuclides in question, several cases are studied. These cases consider for the four analyzed radionuclides inventories produced for three fuel types-pressurized water reactor uranium oxide (PWR UOX), pressurized water reactor mixed oxide (PWR MOX), and advanced high-temperature gascooled reactor (AHTGR)-several burnup values and time out of reactor extending to 200 y. Doses to the maximum exposed individual (MEI) are calculated with the EPA code CAP-88 (Rosnick 2007(Rosnick , 1992. Two dose cases are considered. The first case, perhaps unrealistic, assumes that all of the allowable dose is assigned to the volatile radionuclides. In lieu of this, for the second case a value of 10% of the allowable dose is arbitrarily selected to be assigned to the volatile radionuclides. The required decontamination factors (DFs) are calculated for both of these cases, including the case for the thyroid dose for which 14 C and 129 I are the main contributors. However, for completeness, for one fuel type and burnup, additional cases are provided, allowing 25% and 50% of the allowable dose to be assigned to the volatile radionuclides. Because 3 H and 85 Kr have relatively short half-lives, 12.3 y and 10.7 y, respectively, the dose decreases with the time from when the fuel is removed from the reactor and the time it is processed (herein "fuel age"). One possible strategy for limiting the discharges of these short half-life radionuclides is to allow the fuel to age to take advantage of radioactive decay. Therefore, the doses and required DFs are calculated as a function of fuel age. Here we calculate, given the above constraints and assumptions, the minimum ages for each fuel type that would not require additional effluent controls for the shorter half-life volatile radionuclides based on dose considerations. With respect to 129
The Idaho Nuclear Technology and Engineering Center (INTEC) was home to nuclear fuel reprocessing activities for decades at the Idaho National Engineering and Environmental Laboratory. As a result of the reprocessing activities, INTEC has accumulated approximately one million gallons of acidic, radioactive, sodium-bearing waste (SBW). The purpose of this demonstration was to investigate a steam reforming technology, offered by THOR sm Treatment Technologies, LLC, for treatment of the SBW into a "road ready" waste form that would meet the waste acceptance criteria for the Waste Isolation Pilot Plant (WIPP). A non-radioactive simulated SBW was used based on the known composition of waste tank WM-180 at INTEC. Rhenium was included as a nonradioactive surrogate for technetium.Data were collected to determine the nature and characteristics of the product, the operability of the technology, the composition of the off-gases, and the fate of key radionuclides (cesium and technetium) and volatile mercury compounds. The product contained a significant fraction of elemental carbon residues in the cyclone and filter vessel catches. Mercury was quantitatively stripped from the product but cesium, rhenium (Tc surrogate), and the heavy metals were retained. Nitrates were not detected in the product and NO x destruction exceeded 98%. The demonstration was successful. iv v SUMMARY THOR sm Treatment Technologies, LLC (TTT) was awarded a contract to demonstrate its steam reforming technology on non-radioactive, simulated tank WM-180 sodium-bearing waste using government furnished equipment built and operated by Science Applications International Corporation (SAIC) in Idaho Falls, Idaho. TTT specified the flow sheet conditions and provided additives for the demonstration. Performance dates were January 6 through January 26, 2003 to conduct preliminary optimization tests and execute a successful 100-hour demonstration run.After a few days of proving and optimizing the flow sheet conditions, the demonstration run was started January 13 and completed January 17, 2003. The 100-hr demonstration run was successfully completed. The sodium-bearing waste simulant was converted into a freely-flowing powder and NO x destruction was excellent. Details of the process flow sheet and data that were collected on product and off-gas characteristics are contained within the report. vi vii ACKNOWLEDGMENTSThe demonstration test is a product of diligent efforts from many persons in several different organizations. Test system design and construction, and test operation, was funded by the U.S. Department of Energy through the Idaho National Engineering and Environmental Laboratory (INEEL) High Level Waste Program Idaho Tank Farm Project. The INEEL designed and fabricated the reformer vessel, provided a high-level design for the complete process, and directed the test system installation at the Science Applications International Corporation's (SAIC) Science and Technology Application Research (STAR) Center in Idaho Falls, Idaho. SAIC completed th...
Four radionuclides have been identified as being sufficiently volatile in the reprocessing of nuclear fuel that their gaseous release needs to be controlled to meet U.S. regulatory requirements (Jubin et al. 2011, 2012). These radionuclides are 3 H, 14 C, 85 Kr, and 129 I. Of these, 129 I has the longest half-life and potentially highest biological impact. Accordingly, control of the release of 129 I is most critical with respect to U.S. regulations for the release of radioactive material in stack emissions. Current U.S. Environmental Protection Agency regulation governing nuclear facilities (40 CFR 190) states that the total quantity of radioactive materials entering the general environment from the entire uranium fuel cycle, per gigawatt-year of electrical energy produced by the fuel cycle, must contain less than 5 mCi of 129 I.
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