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

US regulations 40 CFR 61, 40 CFR 190, and 10 CFR 20 govern release limits for volatile radionuclides contained in the gaseous effluents of used nuclear fuel (UNF) reprocessing facilities. Of the four volatile radionuclides that are restricted by the release limits ( 3 H, 14 C, 85 Kr, and 129 I) 129 I will require the greatest degree of abatement. Previous studies have shown that overall plant decontamination factor (DF) for 129 I must, at minimum, exceed 1,000 to meet regulatory requirements. Iodine-containing off-gas will be present in the dissolver off-gas, the cell off-gas, the vessel off-gas (VOG), the waste off-gas, and the shear off-gas. The VOG will most likely contain 1%-5% of the total iodine at part per billion (ppb) concentrations. A number of studies have examined iodine abatement from the dissolver off-gas, which contains greater than 95% of the iodine inventory of the plant, but very few have examined the recovery of iodine from the VOG stream.

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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.