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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

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A concerted effort over the past few years has been focused on enhancing the core model for the High Flux Isotope Reactor (HFIR), as part of a comprehensive study for HFIR conversion from high-enriched uranium to low-enriched uranium (LEU) fuel. At this time, the core model used to perform analyses in support of HFIR operation is a Monte Carlo model with the MCNP code for the beginning of Cycle 400, which was documented in detail in a 2005 technical report. An updated HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed to serve as reference for the design of an LEU fuel for HFIR.The recent enhancements in modeling and simulations for HFIR that are discussed in the present report include: (1) revision of the 2005 MCNP model for the beginning of Cycle 400 to improve the modeling data and assumptions as necessary based on appropriate primary reference sources-HFIR drawings and reports; (2) improvement of the fuel region model, including an explicit representation for the involute fuel plate geometry that is specific to HFIR fuel; and (3) revision of the Monte Carlo-based depletion model for HFIR, in use since 2009 but never documented in detail, with the development of a new depletion model for the HFIR explicit fuel plate representation.The new HFIR models for Cycle 400 are used to determine various metrics of relevance to reactor performance and safety assessments. The calculated metrics are compared, where possible, with measurement data from preconstruction critical experiments at HFIR, data included in the current HFIR safety analysis report, and/or data from previous calculations performed with different methods or codes. The results of the analyses show that the models presented in this report provide a robust and reliable basis for HFIR analyses.

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