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At least 829 records · Page 46
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Properties of selected non-pertechnetate species relevant to Hanford Tank Waste - 15019

Among radioactive constituents present in the Hanford tank waste technetium-99 (Tc) poses a multitude of problems associated with its radiotoxicity, unique and relatively poorly understood chemistry, and high mobility in the environment. Technetium is one of the most difficult contaminants to be addressed at the U.S. Department of Energy (DOE) Hanford Site because of its complex chemical behavior in tank waste and limited incorporation in the glass waste forms during mid- to high-temperature immobilization processes. The majority of Tc exists in the supernatant and salt cake fractions of both single- and double-shell tanks as pertechnetate TcO{sub 4}{sup -} (oxidation state +7) and in the reduced forms (oxidation state < +7) collectively known as non-pertechnetate (non-TcO{sub 4}{sup -}) species. Our research program is aimed at gaining understanding of the behavior of non-pertechnetate species in various oxidation states including such phenomena as redox speciation, hydrolysis, oligomerization, coordination of small organic ligands present in the tank supernatant, and solubility of the complex species. This paper describes our current results on the characterization and properties of the non-pertechnetate species. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Probabilistic Performance Assessment for Deep Borehole Disposal of Cs/Sr Capsules - 17143

The US Department of Energy (DOE) performed an assessment of disposal options for spent nuclear fuel (SNF) and high-level radioactive waste (HLW) that recommended the consideration of deep borehole disposal (DBD) of smaller DOE-managed waste forms, such as cesium (Cs) and strontium (Sr) capsules. To further assess the safety and viability of the DBD concept, post-closure performance assessment (PA) analyses were performed for DBD of Cs/Sr capsules. The post-closure PA included a reference design concept for Cs/Sr capsule disposal, feature, event, and process (FEP) analysis, scenario development, and modeling using the PFLOTRAN code. PA simulations were run for both nominal and disturbed scenarios for long-term radionuclide transport away from the deep disposal borehole. The nominal (expected) post-closure release scenario included FEPs for short-duration (a few hundred years) thermally-induced upward fluid flux through the borehole seals and/or disturbed rock zone (DRZ) followed by longer-term slow diffusive transport. Simulations included a baseline deterministic run and a set of probabilistic realizations (using Latin Hypercube Sampling (LHS) from parameter distributions) to examine the sensitivity and importance of the long-term radionuclide transport to processes and parameters such as: waste package durability; seal porosity and permeability, disturbed rock zone (DRZ) porosity and permeability; and radionuclide sorption in the seal materials, DRZ, and crystalline host rock. The disturbed scenario included a waste package 'stuck' in the crystalline basement above the emplacement zone near a hypothetical borehole-intersecting fracture. Simulations included two deterministic runs to examine the sensitivity to the regional head gradient. The nominal and disturbed scenario PA results suggest that a favorable safety case can be developed for DBD of Cs/Sr capsules and helped to identify key areas of uncertainty, which in turn can inform future DBD research and development (R and D) activities, including a field demonstration project. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Technetium Incorporation into C14 and C15 Laves Intermetallic Phases

Laves-type intermetallics have been observed to be the dominant phases in a series of alloy compositions being designed for the immobilization technetium in a metallic waste form. The dominant metals in the alloy compositions were Fe-Mo and Fe-Mo-Zr. Alloy composition, Fe-Mo-Zr, also contained Pd, Zr, Cr, and Ni. Both non-radioactive rhenium-containing and radioactive technetium-bearing alloy compositions were investigated. In the Fe-Mo series, phases were observed Fe2Mo (C14 Laves phase) and ferrite in agreement with predictions. Both Tc and Re resided predominantly in the Laves phase. In the Fe-Mo-Zr system, the phases included hexagonal C14 with the composition (Fe,Cr)2Mo, cubic C15 phase with a (Fe,Ni)2Zr composition, and the hcp phase Pd2Zr.

alloy↗
DOE OSTI2011

Deep Bed Iodine Sorbent Testing FY 2011 Report

Nuclear fission results in the production of fission products (FPs) and activation products that increasingly interfere with the fission process as their concentrations increase. Some of these fission and activation products tend to evolve in gaseous species during used nuclear fuel reprocessing. Analyses have shown that I129, due to its radioactivity, high potential mobility in the environment, and high longevity (half life of 15.7 million years), can require control efficiencies of up to 1,000x or higher to meet regulatory emission limits. Deep-bed iodine sorption testing has been done to evaluate the performance of solid sorbents for capturing iodine in off-gas streams from nuclear fuel reprocessing plants. The objectives of the FY 2011 deep bed iodine sorbent testing are: (1) Evaluate sorbents for iodine capture under various conditions of gas compositions and operating temperature (determine sorption efficiencies, capacities, and mass transfer zone depths); and (2) Generate data for dynamic iodine sorption modeling. Three tests performed this fiscal year on silver zeolite light phase (AgZ-LP) sorbent are reported here. Additional tests are still in progress and can be reported in a revision of this report or a future report. Testing was somewhat delayed and limited this year due to initial activities to address some questions of prior testing, and due to a period of maintenance for the on-line GC. Each test consisted of (a) flowing a synthetic blend of gases designed to be similar to an aqueous dissolver off-gas stream over the sorbent contained in three separate bed segments in series, (b) measuring each bed inlet and outlet gas concentrations of iodine and methyl iodide (the two surrogates of iodine gas species considered most representative of iodine species expected in dissolver off-gas), (c) operating for a long enough time to achieve breakthrough of the iodine species from at least one (preferably the first two) bed segments, and (d) post-test purging with pure N 2 to drive loosely or physisorbed iodine species off of the sorbent. Post-test calculations determine the control efficiencies for each bed, iodine loadings on the sorbent, and mass transfer zone depths. Portions of the iodine-laden sorbent from the first bed of two of the tests have been shipped to SNL for waste form studies. Over the past three years, we have explored a full range of inlet iodine and methyl iodide concentrations ranging from ~100 ppb to ~100 ppm levels, and shown adequate control efficiencies within a bed depth as shallow as 2 inches for lower concentrations and 4 inches for higher concentrations, for the AgZ-type sorbents. We are now performing a limited number of tests in the NC-77 sorbent from SNL. Then we plan to continue to (a) fill in data gaps needed for isotherms and dynamic sorbent modeling, and (b) test the performance of additional sorbents under development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Experimental Verification of Solidification Stress Theory

A research program is being conducted to develop a crack-free ceramic waste form (CWF) to be used for long term encasement of fission products and actinides resulting from processing spent nuclear fuel. Cracking usually occurs in the cooldown phase of the glass or ceramic formations. A crack-free formation should have more resistance to leaching than one with many cracks. In the research leading up to producing a CWF, a model was developed that proposes a permanent stress develops when the melt solidifies and that this stress can cause failure as the CWF nears room temperature. This paper reports on how the formation, CWF2, confirms the existence of this stress. The solidification stress is in addition to and of opposite sign of the thermal stress. Its derivation is reported on in Ref. 1. Cracking of the CWF would occur at low temperatures if solidification stress exists but at high temperatures if it doesn’t. If solidification stress occurs, then the cooldown rate during solidification should be reduced. If not, it should be reduced when the thermal stresses are highest. Recording cracking sounds confirm the existence of this solidification stress since cracking occurred during the low temperature phase of the cooldown. As a side purpose of this paper, a cooldown rate is proposed that should eliminate cracking in the next experiment, CWF3. CWF2 is a prototype vertical ceramic waste cylinder formed over a period of 10 days by heating a mixture of 75% zeolite, 25% glass frit in an argon atmosphere furnace through melting to 925 C and then cooling through solidification to room temperature. It is approximately 1 m high, 0.5 m in diameter, weighs about 400 kg, and is formed in a stainless steel can 0.5 cm thick. This cylinder developed many cracks on cooldown. At least 15 loud cracks were recorded over a period of 4 days at the end of cooldown when the temperatures were below 400 C. The CWF2 surface and centerline temperatures at mid height were measured which allowed the stress to be calculated. The timing of the cracks was compared to the time the calculated total stress exceeded the tensile stress limit and verified that the cause of the cracking was solidification stress and not thermal stress. Since the CWF is encased in a can in a furnace, the cracks cannot be easily observed but can be detected with sound measurements. Similarly, the stress cannot be measured but only estimated with analysis. Destructive examination of the CWF after cooldown was used to determine the pattern of the cracking which appeared to be initiated mainly in the inner region which is evidence the cracking is due to solidification stress.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Fragmentation Prediction for a Brittle Ceramic: a Two-Scale Model Approach

Glass and oxide-ceramic radioactive material forms (e.g., PuO{sub 2}) are susceptible to brittle fracture during handling and transportation. Fracture of these materials may result in the generation of fragments small enough to become airborne, and thus easily dispersed and inhaled. Entrainment and dispersal of particles with dimensions less than 10 μm is generally anticipated under normal aerodynamic conditions. The assessment of the risk to the public and the environment from these materials due to events resulting in their fracture is based on the determination of the source term for environmental release. This source term is typically quantified by an airborne release fraction (ARF) and a respirable fraction (RF) of material generated by a given event. While the recommended empirical bounding values prove satisfactory to comply with the regulatory process for handling and transportation of these materials, they are based on test data for only a few materials subjected to a limited range of loading conditions. Because the final state of fragmentation (in terms of the size and distribution of fragments) is strongly dependent not only on the properties of the material and the energetic loading conditions that lead to breakup, but also on the many intrinsic length scales associated with fragmentation, it is difficult to precisely predict fragmentation characteristics for other brittle solid wastes forms subjected to a wider range of loading conditions based on the test data alone. The purpose of this manuscript is to present a two-scale modeling approach that may be used to simulate the dynamic fragmentation of brittle solid radioactive material forms subjected to general loading and boundary conditions. The method provides the resulting fragment characteristics (average size and size distribution) across the entire fragment range of interest. At the macro-scale, the approach uses a gradient damage elastic continuum mechanics material model within a finite element framework to simulate dynamic fracture under general loading and boundary conditions. Results from the macroscopic calculations are passed to a one-dimensional (1-D) lower length scale model that incorporates the relevant fine scale effects to determine the micro-scale fragmentation. This 1-D micro-scale fragmentation model simulates the internal crack nucleation and opening processes to predict the average fragment size and the fragment size distribution as a function of the material properties, loading conditions, and microscopic characteristic length scale of the material. Results from the lower length scale model are then combined with the macro-scale model fragmentation information to determine the resulting fragment size distribution spanning both length scales. This two-scale model has been exercised on a representative problem drawn from experimental studies, consisting of an impact of a falling weight onto a brittle ceramic. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2012

Dissolution Behavior of Plutonium Containing Zirconia-Magnesia Ceramics

This study explores the dissolution properties of zirconia-magnesia ceramics containing plutonium as the basis of an inert atrix nuclear fuel. The magnesium oxide phase remains pure MgO, while the zirconia incorporates a small amount of magnesium oxide along with all of the plutonium oxide and erbium oxide. The performance of the material under reactor and repository environments was examined. Reactor conditions are examined using a pressure vessel to expose the material to 300 degrees C water. To assess the performance of the material as a waste form it was submerged in 90 degrees C water for 1000 h. In both aqueous dissolution studies there was minimal release of less than 0.8 wt.% of plutonium from the material. To examine the potential for recycling, the dissolution behavior of the fuel matrix was examined in acidic solutions: pure nitric acid and a nitric acid-hydrofluoric acid-peroxide solution. Both acidic media exhibit potential for dissolving plutonium from the zirconia matrix. The experiments performed in this study are meant to lay a foundation for the chemical performance of zirconia-magnesia inert matrix fuel containing fissile material and burnable poison.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

FUNCTIONALIZED SILICA AEROGELS: ADVANCED MATERIALS TO CAPTURE AND IMMOBILIZE RADIOACTIVE IODINE

To support the future expansion of nuclear energy, an effective method is needed to capture and safely store radiological iodine-129 released during reprocessing of spent nuclear fuel. Various materials have been investigated to capture and immobilize iodine. In most cases, however, the materials that are effective for capturing iodine cannot subsequently be sintered/densified to create a stable composite that could be a viable waste form. We have developed chemically modified, highly porous, silica aerogels that show sorption capacities higher than 440 mg of I2 per gram at 150 C. An iodine uptake test in dry air containing 4.2 ppm of iodine demonstrated no breakthrough after 3.5 h and indicated a decontamination factor in excess of 310. Preliminary densification tests showed that the I2-loaded aerogels retained more than 92 wt% of I2 after thermal sintering with pressure assistance at 1200 C for 30 min. These high capture and retention efficiencies for I2 can be further improved by optimizing the functionalization process and the chemistry as well as the sintering conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

FY-16 Technology Gap Study Technical Report: Analysis of Undissolved Anode Materials of Mark-IV Electrorefiner

The Experimental Breeder Reactor II (EBR-II) is a sodium cooled fast reactor developed at Argonne National Laboratory (ANL). The used fuels from the EBR-II are currently being treated in the Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL). The Mark IV (Mk-IV) electrorefiner (ER) is a unit process in the FCF, which is primarily assigned to treating the used driver fuels. The stainless steel anode baskets hold the chopped spent driver fuel segments. During electrorefining, the anode baskets are immersed into the electrolyte and the used fuel is dissolved electrochemically. Perforated sides and bottoms allow the flow of the electrolyte into and out of the anode baskets. The steel cathode is also immersed into the electrolyte and collects the reduced products. The active metal contents in the used fuel (e.g., Cs, Sr, lanthanides, Pu, etc.) reacts with uranium cations in the electrolyte and progressively reports to the electrolyte. Noble metals are mostly retained in the cladding hulls. Varying quantities of zirconium are retained in the cladding hulls depending on the operational conditions of the Mk-IV ER. The undissolved anode materials are removed from the anode baskets and stored for subsequent metal waste form processing. These undissolved materials typically include undissolved fuels, stainless steel cladding, and adhering electrolyte. A couple of hulls are retrieved for chemical analysis and used for estimating the composition of the entire undissolved anode materials. The mass balance attempt based on this practice of estimating the undissolved anode materials has been a challenge due to inherently high sampling errors associated with heterogeneous undissolved material compositions. Responding to the prescribed challenge, this report investigates chemical analysis data as a whole and finds noticeable trends in the compositions of undissolved anode material samples with respect to the mass of the whole undissolved anode materials. Based upon this discovery, an empirical model is proposed.

36 MATERIALS SCIENCE↗
DOE OSTI2017

The Pretreatment and Immobilization Test on The Greater Than Class C Spent Ion Exchange Resins from Taiwan Research Reactor - 17177

In this work, the experience on the removal of greater than Class C bead ion exchange resins from Taiwan Research Reactor spent fuel pool was reported. All of the spent ion exchange resins had been flushed, dewatered, and packed into lead-shield containers for 60 drums and pending for stabilization. The meta-kaolin type of geo-polymer was selected to immobilize the ion exchange resin. According to the 'Regulations on Final Disposal of Low Level Radioactive Waste and Safety Management of the Facilities' Act in Taiwan, the immobilization waste form shall meet seven testing items. Herein, several immobilize samples were prepared with the following amounts of ion exchange resin, 6, 12 and 18 wt.%. The compressive strength of the samples ranged from 2.9 to 18.2 MPa. The Cs and Sr ions were effectively trapped by ion exchange resins. After the resins incorporated with the geo-polymer matrix, the captured ions were expected difficult to leach out. However, it is found that at the higher contents of ion exchange resins in the immobilize samples exhibited poor water resistance due to the swelling effect. Several parameters such as the ratios of SiO{sub 2}/Na{sub 2}O and powder/solution also the blast furnace slag dosage were investigated to improve processibility and improve performance. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

The DISTINCTIVE University Consortium: Legacy Ponds and Silo Wastes - 18135

The UK Engineering and Physical Sciences Research Council (EPSRC) sponsored DISTINCTIVE consortium (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories) is undertaking 22 science and engineering projects across 11 universities related to the decommissioning of the Legacy Ponds and Silos at Sellafield. These are grouped into three work packages that focus on wasteform durability, effluent treatment and pond and silo sludges. Specific challenges include understanding the durability of intermediate level waste (ILW) glass-ceramic waste forms, studying thermally treated Pu-containing materials, producing and testing new inorganic ion exchange materials including derivatized magnetic nanoparticles for Cs, Sr and U removal from effluent, developing micro- and ultra-filtration methods, providing improved modelling and simulation of sludge disturbance, mobilisation and transport, developing a better understanding of gas hold-up in sludges, exploring methods for sludge characterisation including the use of acoustic backscatter and quartz crystal microbalance techniques and Raman spectroscopy and computer modelling of the interaction of cations with brucite surfaces. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Conditioning of Boron-Containing Low and Intermediate Level Liquid Radioactive Waste - 12041

Improved cementation of low and intermediate level radioactive waste (ILW and LLW) aided by vortex electromagnetic treatment as well as silica addition was investigated. Positive effects including accelerated curing of boron-containing cement waste forms, improve end product quality, decreased product volume and reduced secondary LRW volume from equipment decontamination were established. These results established the possibility of boron-containing LRW cementation without the use of neutralizing alkaline additives that greatly increase the volume of the final product intended for long-term storage (burial). Physical (electromagnetic) treatment in a vortex mixer can change the state of LRW versus chemical treatment. By treating the liquid phase of cement solution only, instead of the whole solution, and using fine powder and nano-particles of ferric oxides instead of separable ferromagnetic cores for the activating agents the positive effect are obtained. VET for 1 to 3 minutes yields boron-containing LRW cemented products of satisfactory quality. Silica addition at 10 % by weight will accelerate curing and solidification and to decrease radionuclide leaching rates from boron-containing cement products. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Criticality safety analysis on fissile materials in Fukushima reactor cores

The present study focuses on the criticality analysis for geological disposal of damaged fuels from Fukushima reactor cores. Starting from the basic understanding of behaviors of plutonium and uranium, a scenario sequence for criticality event is considered. Due to the different mobility of plutonium and uranium in geological formations, the criticality safety is considered in two parts: (1) near-field plutonium system and (2) far-field low enriched uranium (LEU) system. For the near-field plutonium system, a mathematical analysis for pure-solute transport was given, assuming a particular buffer material and waste form configuration. With the transport and decay of plutonium accounted, the critical mass of plutonium was compared with the initial load of a single canister. Our calculation leads us to the conclusion that our system with the initial loading being the average mass of plutonium in an assembly just before the accident is very unlikely to become critical over time. For the far-field LEU system, due to the uncertainties in the geological and geochemical conditions, calculations were made in a parametric space that covers the variation of material compositions and different geometries. Results show that the LEU system could not remain sub-critical within the entire parameter space assumed, although in the iron-rich rock, the neutron multiplicity is significantly reduced.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Formulation and preparation of Hanford Waste Treatment Plant direct feed low activity waste Effluent Management Facility core simulant

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate, LMOGC) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream is to evaporate it in a new evaporator in the Effluent Management Facility (EMF) and then return it to the LAW melter. It is important to understand the composition of the effluents from the melter and new evaporator so that the disposition of these streams can be accurately planned and accommodated. Furthermore, alternate disposition of the LMOGC stream would eliminate recycling of problematic components, and would enable less integrated operation of the LAW melter and the Pretreatment Facilities. Alternate disposition would also eliminate this stream from recycling within WTP when it begins operations and would decrease the LAW vitrification mission duration and quantity of glass waste, amongst the other problems such a recycle stream present. This LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures and are problematic for the glass waste form, such as halides and sulfate. Because this stream will recycle within WTP, these components accumulate in the Melter Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Diverting the stream reduces the halides and sulfate in the recycled Condensate and is a key outcome of this work. This overall program examines the potential treatment and immobilization of this stream to enable alternative disposal. The objective of this task was to formulate and prepare a simulant of the LAW Melter Off-gas Condensate expected during DFLAW operations. That simulant can be used in evaporator testing to predict the composition of the effluents from the Effluent Management Facility (EMF) evaporator to aid in planning for their disposition. This document describes the method used to formulate a simulant of this LAW Melter Off-Gas Condensate stream, which, after pH adjustment, is the feed to the evaporator in the EMF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Physical, Hydraulic, and Transport Properties of Sediments and Engineered Materials Associated with Hanford Immobilized Low-Activity Waste

Current plans for treatment and disposal of immobilized low-activity waste (ILAW) from Hanford’s underground waste storage tanks include vitrification and storage of the glass waste form in a nearsurface disposal facility. This Integrated Disposal Facility (IDF) is located in the 200 East Area of the Hanford Central Plateau. Performance assessment (PA) of the IDF requires numerical modeling of subsurface flow and reactive transport processes over very long periods (thousands of years). The models used to predict facility performance require parameters describing various physical, hydraulic, and transport properties. This report provides updated estimates of physical, hydraulic, and transport properties and parameters for both near- and far-field materials, intended for use in future IDF PA modeling efforts. Previous work on physical and hydraulic property characterization for earlier IDF PA analyses is reviewed and summarized. For near-field materials, portions of this document and parameter estimates are taken from an earlier data package. For far-field materials, a critical review is provided of methodologies used in previous data packages. Alternative methods are described and associated parameters are provided.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Aging and iodine loading of silver-functionalized aerogels

Engineered silver-functionalized silica aerogels are being investigated for their potential application in off-gas treatment at a used nuclear fuel reprocessing facility. Reprocessing will release several key volatile radionuclides, including iodine-129. To achieve regulatory compliance, iodine-129 must be removed from any off-gas stream prior to environmental discharge. Ag{sup 0}-functionalized aerogels have been demonstrated to have high iodine-capture capacity, high porosity, and potential for conversion into a waste form. Capture materials used in off-gas treatment may be exposed to a heated, high-humidity, acidic gas stream for months. Extended exposure to this stream could affect sorbent performance. It was the aim of this study to evaluate what impacts might be observed when Ag{sup 0}-functionalized aerogels prepared at Pacific Northwest National Laboratory were contacted with a dry air stream for up to 6 months and then used to adsorb iodine from a synthetic off-gas stream. Results demonstrate that there is some loss of iodine-capture capacity caused by aging, but that this loss is not as marked as for aging of more traditional iodine sorbents, such as silver-impregnated mordenite. Specifically, aging silver-functionalized aerogel under a dry air stream for up to 6 months can decrease its iodine capacity from 41 wt% to 32 wt%. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Summary Report on the Volatile Radionuclide and Immobilization Research for FY2011 at PNNL

The materials development summarized here is in support of the Waste Forms campaign, Volatile Radionuclide task. Specifically, materials are being developed for the removal and immobilization of iodine and krypton, specifically 129I and 85Kr. During FY 2011, aerogel materials were investigated for removal and immobilization of 129I. Two aerogel formulations were investigated, one based on silica aerogels and the second on chalcogen-based aerogels (i.e., chalcogels). A silica aerogel was tested at ORNL for total I2 sorption capacity. It was determined to have 48 mass% capacity while having little physisorbed I2 (I2 not taken up in the aerogel pores). For 85Kr, metal organic framework (MOF) structures were investigated and a new MOF with about 8 mass% capacity for Xe and Kr. The selectivity can be changed from Xe > Kr to Xe < Kr simply by lowering the temperature below 0 C. A patent disclosure has been filed. Lastly, silicon carbide (SiC) was loaded with Kr. The diffusion of Kr in SiC was found to be less than detectable at 500 C.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Review of recent developments in iodine wasteform production

Radioiodine capture and immobilization is not only important to consider during the operation of reactors (i.e., I-131), during nuclear accidents (i.e., I-131 and I-129) or nuclear fuel reprocessing (i.e., I-131 and I-129), but also during disposal of nuclear wastes (i.e., I-129). Most disposal plans for I-129-containing waste forms (including spent nuclear fuel) propose to store them in underground repositories. Here, iodine can be highly mobile and, given its radiotoxicity, needs to be carefully managed to minimize long-term environmental impacts arising from disposal. Typically, any process that has been used to capture iodine from reprocessing or in a reactor is not suitable for direct disposal, rather conversion into a wasteform for disposal is required. The objectives of these materials are to use either chemical immobilization or physical encapsulation to reduce the leaching of iodine by groundwaters. Some of the more recent ideas have been to design capture materials that better align with disposal concepts, making the industrial processing requirements easier. Research on iodine capture materials and wasteforms has been extensive. This review will act as both an update on the state of the research since the last time it was comprehensively summarized, and an evaluation of the industrial techniques required to create the proposed iodine wasteforms in terms of resulting material chemistry and applicability.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗