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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 757 records · Page 42
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Material Balance Area Design for the Transuranic Fuel Cycle Employing High Temperature Gas Cooled Reactors

A major impetus has been to ensure safety in nuclear power plant systems from the beginning of its development to the current period of nuclear renaissance. As one of the next generation reactors with a high level of nuclear safety, high temperature gas cooled reactor (HTGR) has attracted large attention because of the possibility to shut down the reactor safely and remove decay heat without the use of active cooling. A deep-burn concept of fuel utilization is proposed and studied for HTGRs to alleviate plutonium proliferation concern. The idea is to utilize the plutonium extracted by reprocessing the light water reactor (LWR) spent fuel and discharge it only at higher fuel burn-ups of up to 90 atom %. In addition HTGRs have been studied to utilize transuranic (TRU) fuel. A large amount of special nuclear material (SNM) will be used in the HTGR advanced fuel cycle. The inherent proliferation resistance and material attractiveness of the fresh fuel and discharged spent fuel have been discussed against the diversion from peaceful to military uses. On the other hand safeguards of SNM is also important to prevent nuclear proliferation under the Non-Proliferation Treaty between the state and International Atomic Energy Agency (IAEA). Facility operators are responsible to design material balance area (MBA) and provide enough information on material balance of SNMs for the safeguards conclusion. The material unaccounted for (MUF) and measurement uncertainty in MUF (σ{sub MUF}) represent accuracy of material balance. The MBAs have to be designed and established to achieve enough confidence level on MUF and σ{sub MUF}. The MBA design is challenging for nuclear facilities managing large amount of bulk (instead of items) SNM because of potentially large MUF and σ{sub MUF}. A concept and guidance of MBA design for uranium fueled HTGRs has been discussed. In this study, a MBA design is presented for the transuranic fuel cycle employing HTGRs. The MUF and the σ{sub MUF} is calculated and evaluated using international target values (ITVs) provided by the IAEA. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Neutronics and computational fluid dynamics analyses for a natural circulation loop of molten fuel salt

A conceptual reactor system of molten fuel salt loop designed for Mo-99 production was analyzed by coupling TRITON neutronics and Fluent computational fluid dynamics (CFD) simulations. The loop system operated in 1 MW thermal power features natural circulation and potential extraction of non-soluble fission products (FP) in fuel salt. The behavior of natural circulation was studied and the power and temperature distributions in equilibrium were obtained. A long-term fuel depletion analysis of the system considering fuel circulation and reprocessing was then performed using an approximate stepwise sequence of simulations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

CANDLE Burning Reactor with Plutonium Fuel Start-up Core

The CANDLE burning fast reactor concept makes it possible to utilize natural uranium or depleted uranium with high burnup of up to 40% without repossessing of spent fuel from the reactor. Once the reactor reaches equilibrium burning wave condition, it needs only natural or depleted uranium, although it needs some amount of fissile material for the startup. Studies have been performed for the startup core using enriched uranium and absorbers. It is expected that plutonium obtained by reprocessing of spent fuel from light water reactors (LWR) will be an excellent fuel for the startup core. Using this fuel for the startup core will eliminate the need to produce enriched uranium, as well as make it possible to utilize the plutonium from spent fuel effectively as a useful resource. For effective utilization of plutonium as a resource, a startup core that requires only a small amount of plutonium fuel is preferable. The purpose of this study was to clarify the CANDLE burning fast reactor concept using a minimum amount of plutonium fuel from LWR spent fuel for the startup core by analysis with a simplified reactor model. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Aging Processes of Silver Mordenite and Silver Functionalized Aerogel in Dry Air, Humid Air and NO/N{sub 2}

One major problem in the treatment of nuclear-fuel reprocessing off-gas streams is the loss of capacity of the adsorbents exposed to off-gas streams containing air, I{sub 2} vapor, H{sub 2}O vapor, NO and NO{sub 2} over extended periods of time. These aging processes decrease the adsorption capacity of the reduced silver exchanged mordenite (Ag{sup 0}Z) and silver functionalized silica aerogel (Ag{sup 0}-aerogel) columns during the long processing time of iodine adsorption. Significant work has been performed over the past few years at ORNL on Ag{sup 0}Z and Ag{sup 0}-aerogel aging. The Ag{sup 0}Z and Ag{sup 0}-aerogel were aged in 4 different gas streams at 150 deg. C for periods of time up to 6 months. The aging samples were then loaded with I{sub 2} to determine the capacity loss after the aging. The results are summaries in Tab. 1. It was found that both Ag{sup 0}Z and Ag{sup 0}-aerogel had different amounts of capacity loss in different gases. For example, 40% and 22% capacity loss were observed for Ag{sup 0}Z and Ag{sup 0}-aerogel when aged in dry air for 6 months, while 78% and 43% capacity loss were observed when aged in 1% NO{sub 1} for 2 months. In addition, it was found that Ag{sup 0}Z and Ag{sup 0}-aerogel had a different capacity loss when aged at the same conditions. For example, after aged in 2% NO{sub 2} for 2 months, Ag{sup 0}Z lost 30% iodine capacity, but there was no capacity loss for Ag{sup 0}- aerogel. The differences in capacity loss should be due to different aging processes occurred when Ag{sup 0}Z and Ag{sup 0}- aerogel were exposed to different gases. The mechanisms of the aging processes have not been determined and need to be studied with proper chemical and physical characterization methods. Towards this objective, experimental work is proposed and on-going to obtain data to determine the mechanisms of the aging processes of Ag{sup 0}Z and Ag{sup 0}-aerogel in different aging gas streams. In addition, models will be developed to describe the deactivation process of the adsorbents when exposed to off gases. The deactivation model will be included in the transport models for predicting the adsorption processes over these long-time periods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Nuclear Fuel Cycle Simulator as a Means to Model a Nuclear Hybrid Energy System

With a growing concern for how greenhouse gas (GHG) emissions impact climate change, there is a clear value in minimizing reliance on fossil fuels. Fluctuating renewable energy sources continue to penetrate the grid at higher levels. One means of ensuring a reliable electricity supply, while maintaining low overall GHG emissions, is to couple a nuclear power plant to the fluctuating renewable electricity source. Nuclear hybrid energy systems (NHES) focus on combining renewable and nuclear generation with industrial processes to ensure reliable energy generation as well as profitability of the system as a whole. This paper analyzes the current state of NHES models, and suggests how a nuclear fuel cycle simulator (NFCS) could be used to quickly model a nuclear hybrid energy system to address long term policy concerns. NFCSs are used to model the nuclear fuel cycle, which consists of uranium mining, enrichment, the reactor, reprocessing, and long term storage of waste. The simulations can span over hundreds of years to demonstrate different approaches to managing costs, waste disposal, and proliferation concerns. Current NHES models, on the other hand, focus on shorter time periods, primarily on a daily to weekly scale, and answer questions surrounding feasibility and costs for both normal and extreme scenarios. First, this paper will discuss the different criterion that are important to modeling a NHES. Then, we will break down the characteristics of a NFCS, analyzing how the characteristics of NHES and the fuel cycle overlap. After discussing the current modeling approaches for NHES and NFCS, this paper will analyze the benefits of using a NFCS to model a NHES. This paper will conclude that a NFCS, if appropriately augmented to account for the dynamic aspects of a NHES, can produce a flexible, high level NHES model quickly.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Electrochemical Behavior of Lanthanum in Molten LiF-NaF-KF 'Flinak' Eutectic Salt

In order to develop the Fluoride High- Temperature Reactor, an emerging Gen-IV reactor concept, a primary coolant cleanup system must be developed to allow online reprocessing of the primary molten fluoride salt. This research is focused on the determination of electrochemical properties of rare earths (REs) in molten fluoride salt related to the FHR. The FHR utilizes two fluoride salts. These fluoride salts are LiF-NaF-KF (46.5-11.5-42 mol%), melting point 454 deg. C, also known as 'Flinak', and LiFBeF{sub 2} (33-67 mol%), melting point 459 deg. C, also known as 'Flibe'. Flibe is selected as the primary coolant salt due to its low neutron cross-section (Li{sup 7}-enriched), superior heat transfer properties, and low melting point. However, due to the beryllium content in Flibe, it is highly toxic to humans and Flinak is preferred for the secondary and tertiary coolant salts. The work presented here is specifically focused on the behavior of lanthanum trifluoride (LaF{sub 3}) in molten Flinak. Furthermore, updated data and analysis regarding the electrochemical behavior of dysprosium trifluoride (DyF{sub 3}) is also presented. Previous work related to this topic was presented at the 2016 ANS winter conference. The goal of this research is to develop a contaminant cleanup system for the primary loop of the FHR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

A study on the criticality safety for the direct disposal of used nuclear fuel in Japan: Application of burnup credit to the criticality safety evaluation for the disposal canister - 14049

Japan has recently started the technical development of the direct disposal of used nuclear fuel (UNF) to prepare various disposal options. Unlike the disposal of radioactive waste from UNF reprocessing, criticality safety is important here because of the presence of a certain amount of fissile nuclides in UNF. This paper provides an outline of the research to be addressed in this field and the relevant studies in Japan. It especially includes the first result of a criticality safety evaluation of a disposal canister model adopting burnup credit. In this evaluation, it was assumed that a steel disposal canister loaded with fuel assemblies was surrounded by a clay buffer material, and the fuel assemblies and the disposal canister were intact. Conditions representative of discharged pressurized-water reactor (PWR) fuel were applied in this study - initial U-235 enrichment of 4.5 wt% and burnup of 45 GWd/t. The isotopic composition of the burnt fuel was evaluated using SWAT3.1 code. In the depletion calculation, conservative parameters were applied within the typical PWR operation range. The effective neutron multiplication factor (k{sub eff}) was then calculated using a continuous-energy Monte Carlo code MVP-2.0, adopting the JENDL-4.0 nuclear data library. Uncertainty in k{sub eff} due to depletion calculation errors was evaluated using the correction factors of isotopic composition, which were obtained via multiple comparisons between the calculated and measured isotopic concentrations of irradiated fuels. In addition, the effects of axial burnup profile and horizontal burnup gradient on k{sub eff} were evaluated using the open burnup distribution database. The k{sub eff} including the uncertainties and conservatism was found to be below 0.95 for both actinide-only and actinide and fission product considerations when the fuel assemblies and the disposal canister were assumed to be intact. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

A burnup credit approach irradiated fast-neutron reactor fuels/ nuclides of interest and fuel storage application - 14248

The concept of taking into account the reduction of the reactivity of nuclear used fuel due to their burnup is referred to as 'Burnup Credit' (BUC). Economic incentives offered by the implementation of a BUC methodology in criticality analyses of Light Water Reactor (LWR) used fuel facilities are nowadays well-demonstrated. The current prospect of reprocessing the fissile and fertile subassemblies of the PHENIX reactor in France, and the development of the 4.-Generation Fast-neutron Reactors (FR) enable research projects in relevant fuel cycle operations. Under these circumstances, the CEA and AREVA-NC have decided to study a burnup credit approach for irradiated fuels of the PHENIX reactor. This paper focuses on a preliminary investigation on the use of burnup credit for used FR fuel operations. The first step consists in analyzing and selecting the nuclides of interest for burnup calculation, on the basis of the 27 BUC nuclides (12 actinides and 15 fissions products) chosen for LWR-MOx fuels [1]. These nuclides are then involved in a criticality calculation (a used fuel pool) to highlight the interest of a BUC approach. This study highlights the interest of a burnup credit approach for FR fuels and shows that the highest contribution to reactivity loss is due to only one fission product: the {sup 149}Sm. The opportunity to take account of only one fission product is highly promising and will considerably simplify the way to determine a penalizing fuel inventory. Moreover, it is to notice that results for different configurations may differ from those obtained for a fuel storage application. Thus, specific analyses have to be carried out in order to study the applicability of the conclusions of this paper. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

SFDS: The criticality assembly for the experimental research of the nuclear spent fuel dissolver in CIAE - 14401

SFDS is the criticality assembly for the experimental research of the spent fuel dissolver in China Institute of Atomic Energy. The spent fuel dissolver, as the bottleneck of spent fuel reprocessing, has so many characteristics such as the dynamic changes, heterogeneous distribution, complex compositions that it is hard to analysis by theoretical method. Without enough experiments, clear bias and uncertainty of the analysis tools the handling limits for criticality safety in dissolver can't be determined. The SFDS, physical designed by MONK 9A code, uses two kinds of nuclear fuels, 3% enrichment solid UO{sub 2} rods and 5.5% enrichment UO{sub 2}(NO{sub 3}){sub 2} solution, will meet the demand of the critical experiments about the dissolvers. Seven series of critical experiments will be present on it, those are on pitches of rods, uranium concentrations of solutions, nitric acid concentrations of solutions, fission products, void fractions, temperatures and dissolve phases. More kinds of experiments could be introduced with the flexible structures of SFDS in the future. The current work on it are testing and debugging with full systems by water and it will be first critical later in 2015. It will become the first benchmark assembly for critical experiments research in the nuclear fuel cycle system in China. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Experimental validation of the Cristal V2.0 French criticality calculation package - 14587

A new version of the CRISTAL criticality calculation package, named CRISTAL V2.0 [1], has been developed by IRSN and CEA, in collaboration with AREVA and EDF. Four calculation routes are now available, using nuclear data issued from the JEFF-3.1.1 evaluation: two multigroup routes based on 281 groups cross-sections (APOLLO2 - MORET 5 or APOLLO2-Sn calculations), a pointwise Monte Carlo route (TRIPOLI-4{sup R}) and a criticality standard calculation route. The CRISTAL V2.0 package, which will be delivered soon to OECD/NEA data bank, benefits from a broad validation database covering almost all areas of criticality-safety applications, from fuel fabrication to reprocessing, including transportation. 3,127 critical experiments, corresponding to 342 series, either issued from the OECD/ICSBEP Handbook or performed in French facilities (confidential data) were selected for the CRISTAL V2.0 validation database. Currently, more than 2,300 experiments are already part of the validation database. This validation work highlights an overall consistency to experimental data and inter-comparisons between the different calculation routes mainly show good agreements on a large set of common experiments. (authors)

42 ENGINEERING↗
DOE OSTI2015

Recycling facility periodic safety review: Application of a methodology used for the nuclear criticality safety review of geometrically safe or favorable equipment - 14610

In France, for all fuel cycle facilities, the reference safety frame is continuously adapted to take into account technical developments such as changes in process and/or maintenance operations. Every ten years according to French law, the facility operator is required to launch a process of Periodic Safety Review (PSR), whose conclusions are sent to the Nuclear Safety Authority (NSA). This PSR is split in two parts: first, a review of the facility to its safety documentation and authorization, second, a safety reassessment. Actually, this PSR takes account for topics (actual aging for instance) which require several years of feedback from experience to be integrated in safety evaluations, or which cannot be done on a daily basis (documentation reorganization). It also incorporates the new regulatory safety requirements. A specific methodology was developed for the first PSR of UP3-A since the publication of the law. UP3-A is a used fuel reprocessing facility of La Hague plant. This methodology, adapted for large plants with numerous process and fissile media, is based on a two-step approach intending to focus the analysis on the real hard spots. First, 'sensitive' systems are identified. Second, a specific review is performed on these systems. After a reminder of the methodology, this paper presents the results of its application on the UP3-A scope. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Overview of neutron absorber materials used in spent fuel pools - 14622

Spent fuel pools (SFPs) were originally designed with low density storage racks that maintained subcriticality based on geometric separation. With the increasing need for storage, due to elimination of reprocessing as a viable alternative, SFPs with high-density racks required neutron absorber materials to control reactivity and maintain criticality safety margins. The Nuclear Energy Institute (NEI) conducted an industry wide survey in 2013 and the survey was updated recently based on additional utility data provided by Electric Power Research Institute (EPRI). The survey results show that about one-quarter of the pools in the United States either do not have neutron absorber material in the SFP or do not credit it even when it is present. Also, about half of the plants in the U.S. use BORAL{sup R} as the neutron absorber material followed by Metamic, Carborundum, Alcan, and borated stainless steel. For neutron absorber monitoring, the majority of the plants rely on a coupon testing program. However, for those plants that never installed coupons or have exhausted their supply of coupons from coupon sampling, in situ measurements at periodic intervals can be performed to determine the status of the neutron absorber. In this paper, an overview of the types of neutron absorber materials used in SFPs, the distribution of absorber material use by material type, types of monitoring programs as part of aging management programs, operational experience to date with different absorber materials based on monitoring programs, and industry initiatives toward addressing long-term performance issues related to neutron absorber material degradation and monitoring are presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Backend Nuclear Fuel Cycle Radiation Chemistry

Given global commitments to significantly increase nuclear energy capacity, it is now more important than ever to develop efficient used nuclear fuel management strategies to improve resource utilization, energy security, and waste minimization. Here, an overview of nuclear energy, backend fuel cycle challenges, and advances in used nuclear fuel reprocessing radiation chemistry will be presented. More specifically, the use of electron pulse irradiation techniques to explore radiation-induced reaction mechanisms in actinide containing solutions and solvent systems.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2026

Non-Equilibrium Actinide Radiation Chemistry and the Nuclear Fuel Cycle

Invited John and Naomi Fackler Lectureship in Chemistry and English seminar at Valparaiso University, IN, USA. Actinides are inherently unstable elements that frequently coexist with other radioisotopes, generating intense ionizing radiation fields that drive the formation of non-equilibrium oxidation states. These transient species exert a profound mechanistic influence on the radiation response of actinide-containing systems due to their unique redox chemistry. Despite their importance, they remain poorly understood, yet such insight is essential for advancing actinide science and accurately predicting radiation-driven behavior. Actinide separations—critical for nuclear energy technologies, strategic deterrence, space exploration, and nuclear medicine—depend on precise control of actinide oxidation states to recover targeted elements from complex matrices such as used nuclear fuel. However, during these processes, actinides, their coordination complexes, and the separation media are all exposed to intense, multicomponent (alpha, beta, gamma, etc.) radiation fields that can alter process efficiency, selectivity, and chemical stability. Understanding, controlling, and mitigating radiation-induced reactions is therefore key to innovating and optimizing next-generation separation technologies. This seminar will provide an overview of the nuclear fuel cycle and non-equilibrium actinide radiation chemistry in the context of recovering actinides from used nuclear fuel, with a particular emphasis on direct-dissolution–based reprocessing strategies. We will explore time-resolved electron pulse radiolysis and gamma dose accumulation studies to elucidate the molecular-level roles of radiation-driven, non-equilibrium actinide species in process performance and in the radiolytic stability of organic ligands used for actinide recovery. These insights offer new pathways for designing advanced separation methods and next-generation solvent systems, with broad implications for the future of the nuclear fuel cycle.

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗
DOE OSTI2022

Rare Earth and Transition Metal Containing Glasses

Transition metal (TM) and rare earth (RE) ions have been incorporated into many glass systems such as silicate, phosphate, and borosilicate-based oxide glasses, as well as in halide and chalcogenide glasses, that find applications ranging from optical, photonic, and magnetic devices, solid-state battery, to nuclear waste disposal. Understanding the structural role of RE and TM in these glasses can help to develop glass compositions for targeted applications with either high-optical emission efficiency, electrical conductivity, or chemical durability. In this chapter, we first provide a general introduction of the applications and structural features of RE and TM in glasses, then the critical aspects of molecular dynamics (MD) simulations of these glasses such as interatomic potentials, structural analysis tools to study RE and TM ions in glasses and their clustering behaviors, Quantitative Structure–Property Analysis (QSPR), diffusion and dynamic property calculations, and electronic structure calculations to understand electronic defects such as charge trapping and radiation effects are introduced. Three representative case studies are presented: the first one is on MD simulations of erbium- and europium-doped silica and silicate glasses, as well as cerium doped aluminophosphate glasses, that revealed the effect of glass composition on RE ion local structure and clustering behavior. Electronic structure calculations of cerium-doped glass show how the existence of multioxidation states help to mediate radiation-induced damages caused by excited electron–hole pairs was also discussed. The second one focuses on alkali vanadophosphate glasses where the existence of two vanadium oxidation states help to provide electronic conduction in the glasses while alkali ions provide ionic conduction. MD simulations were used to understand vanadium environments and other structural aspects in the phosphate glasses, as well as the ionic transport behaviors of alkali ions. The third case study is on zirconium-containing borosilicate and aluminosilicate glasses that find wide applications in nuclear waste disposal. MD simulations help to provide structural details of zirconium ions that are validated by diffraction and EXAFS spectra. The structural information was used to interpret changes of mechanical properties and chemical durability by using QSPR and other analyses-based MD-generated structure models.

Du, Jincheng↗
DOE OSTI2023

Fluoride-Catalyzed Siloxane Exchange as a Robust Dynamic Chemistry for High-Performance Vitrimers

Sustainable development of new technologies requires materials having advanced physical and chemical properties while maintaining reprocessability and recyclability. Vitrimers are designed for this purpose; however, their dynamic covalent chemistries often have drawbacks or are limited to specialized polymers. Here, fluoride-catalyzed siloxane exchange is reported as an exceptionally robust chemistry for scalable production of high-performance vitrimers through industrial processing of commodity polymers such as poly(methyl methacrylate), polyethylene, and polypropylene. The vitrimers show improved resistance to creep, heat, oxidation, and hydrolysis, while maintaining excellent melt flow for processing and recycling. Furthermore, the siloxane exchange between different vitrimers during mechanical blending results in self-compatibilized blends without any compatibilizers. In conclusion, this offers a general, scalable method for producing sustainable high-performance vitrimers and a new strategy for recycling mixed plastic wastes.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Demonstration of Complete Recycling Processes of Reversible Epoxies Using Solar Energy Conversion

Reversible epoxies using the Diels–Alder chemistry enables recycling processes through depolymerizing the polymer at higher temperature and then repolymerizing upon cooling. Compared to conventional bulk heating, photothermal heating can save time and resource and, consequently, reduce costs to reach an elevated temperature for recycling processes of the reversible epoxies. In previous studies, self‐healing of cracks and reattachments of two broken pieces have been presented using a laser; however, recycling of a sample as a whole is not feasible by using such a point light source. Herein, complete recycling processes are demonstrated utilizing an area light source, i.e., sunlight. Reversible epoxies are incorporated with carbon black and refractory plasmonic titanium nitride nanoparticles (NPs). Under concentrated (10 times) sunlight, they can generate sufficient heat (≈140 °C) to completely liquefy, reprocess, and reshape the samples multiple times. Recycling processes are validated by evaluation of mechanical properties for each cycle. Using an integrated experimental and theoretical approach, photothermal performance is investigated in terms of the dispersion and loading of photothermal NPs in the matrix, as well as the sample thickness. In this study, an insight is provided into the design of polymer/photothermal nanomaterial composites which can be sustainably recycled using abundant solar energy.

14 SOLAR ENERGY↗
DOE OSTI2016

Adsorption of iodine on hydrogen-reduced silver-exchanged mordenite: Experiments and modeling

The adsorption process of iodine, a major volatile radionuclide in the off-gas streams of spent nuclear fuel reprocessing, on hydrogen-reduced silver-exchanged mordenite (Ag 0 Z) was studied at the micro-scale. The gas-solid mass transfer and reaction involved in the adsorption process were investigated and evaluated with appropriate models. Optimal conditions for reducing the silver-exchanged mordenite (AgZ) in a hydrogen stream were determined. Kinetic and equilibrium data of iodine adsorption on Ag 0 Z were obtained by performing single-layer adsorption experiments with experimental systems of high precision at 373–473 K over various iodine concentrations. Results indicate approximately 91% to 97% of the iodine adsorption was through the silver-iodine reaction. The effect of temperature on the iodine loading capacity of Ag 0 Z was discussed. In conclusion, the Shrinking Core model describes the data well, and the primary rate controlling mechanisms were macro-pore diffusion and silver-iodine reaction. © 2016 American Institute of Chemical Engineers AIChE J, 2016

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗