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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 OSTI2018

The DISTINCTIVE University Consortium: Theme 2: PuO{sub 2} and Fuel Residues - 18132

The safe and secure management of Pu is a matter of serious international concern, with ∼250 t of separated Pu currently stockpiled worldwide. Nearly 50% of this material is in long-term storage in the UK whilst the British government develops its options for final treatment and disposition. Options for this are: (i) reuse as a fuel in modern reactors (e.g. as MOX fuel or within a fast reactor) prior to disposal; (ii) prompt immobilisation for disposal as soon as is practicable; or, (iii) continued long term storage (prior to disposition). The focus of the work presented here is options (ii) and (iii). The R and D needs of both are now pressing: in option (iii) due to it being the current default; in option (ii) because of a comparative lack of R and D on Pu conditioning and packaging due to policy uncertainty as to whether it would be disposed of in a Geologic Disposal Facility (GDF). Addressing these needs is complicated by Pu's high radioactivity, decay heat and radiotoxicity, nuclear safeguard requirements and, for some UK Pu contaminated materials targeted for disposal, poor inventory. Thus, there is also a critical requirement for underpinning research on characterisation methods for Pu bearing materials. In response to these needs, the DISTINCTIVE Consortium (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories, a multi-disciplinary collaboration of 11 universities and 3 key industry partners from across the UK's civil nuclear sector) has been seeking to: - Understand how the structure and properties of PuO{sub 2} change with time in the presence of H{sub 2}O. - Understand the roles these processes play in gaseous product evolution from PuO{sub 2} in storage. - Understand radiation-induced amorphization and dissolution kinetics of Pu waste forms. - Develop novel, fast neutron based radiometric methods for the quantification, isotopic composition assessment and remote imaging of Pu-bearing materials. This paper describes recent progress made towards achieving these objectives. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Compositional effects on the chemical durabilities of aluminophosphate glasses: A review

Phosphate glasses have a range of applications including as hosts for immobilizing radioactive wastes. Studies have shown that addition of iron and/or aluminum oxides can drastically improve the chemical durability of phosphate glasses where the accurate measurement of chemical durability is one of the most important factors for determining the long-term viability of a given waste form. However, due to inconsistencies with the experimental methods used to generate chemical durability data, comparing and interpreting such data is a tedious task. These variables include the temperature of the test, the specimen form (e.g., coupon, particles), the pressure of the test (e.g., atmospheric pressure, elevated pressure in an autoclave), the exposure time, the exposure medium, and how the loss is documented (e.g., total mass lost, normalized elemental release). This review paper summarizes a large collection of chemical durability tests on aluminophosphate glasses in various studies. In addition, the effects of different oxides on the properties of phosphate glasses are summarized.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Corrosion Testing of Monofrax K-3 Refractory in Defense Waste Processing Facility (DWPF) Alternate Reductant Feeds

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) uses a combination of reductants and oxidants while converting high level waste (HLW) to a borosilicate waste form. A reducing flowsheet is maintained to retain radionuclides in their reduced oxidation states which promotes their incorporation into borosilicate glass. For the last 20 years of processing, the DWPF has used formic acid as the main reductant and nitric acid as the main oxidant. During reaction in the Chemical Process Cell (CPC), formate and formic acid release measurably significant H 2 gas which requires monitoring of certain vessel’s vapor spaces. A switch to a nitric acid-glycolic acid (NG) flowsheet from the nitric-formic (NF) flowsheet is desired as the NG flowsheet releases considerably less H 2 gas upon decomposition. This would greatly simplify DWPF processing from a safety standpoint as close monitoring of the H 2 gas concentration could become less critical. In terms of the waste glass melter vapor space flammability, the switch from the NF flowsheet to the NG flowsheet showed a reduction of H 2 gas production from the vitrification process as well. Due to the positive impact of the switch to glycolic acid determined on the flammability issues, evaluation of the other impacts of glycolic acid on the facility must be examined.

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

Enhanced crevice corrosion of stainless steel 316 by degradation of Cr-containing hollandite crevice former

This study explores the corrosion interactions between stainless steel 316 (SS316), a candidate nuclear waste canister material, and Cr-containing hollandite (Ba 1.15 Cr 2.3 Ti 5.7 O 16 ), a model ceramic nuclear waste form. The two materials were corroded in close proximity in 0.6 M NaCl solution at 90 °C for 28 days. Severe crevice corrosion was observed on the stainless steel surface after the corrosion experiment, which is likely associated with the rapid accumulation of Cr 3+ cations originating from both SS316 and the Cr-containing hollandite. Furthermore, the enhanced corrosion of SS316 by the corrosion products of hollandite is quantitatively explained using a classical crevice corrosion model.

36 MATERIALS SCIENCE↗
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 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 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 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↗
DOE OSTI2017

Near-Complete Retention of Technetium During Vitrification of Hanford Low Activity Waste With Recycle - 17315

Technetium is one of the more volatile radionuclides for which retention in low activity waste (LAW) glass under the high temperature conditions during vitrification can vary depending on feed composition, feed chemistry, and melter operating parameters. High retention of technetium in the glass is desirable in order to minimize the fraction of technetium that is directed to secondary waste streams, which include liquid off-gas treatment system effluents. Consequently, test data are crucial to assess and underpin the assumptions underlying such mass balance projections, the compositions of secondary waste streams, and, in particular, the effects of recycle on increasing the incorporation of technetium in the glass products and minimizing the fraction in non-glass waste forms. Previous testing performed on a scaled melter system that incorporates prototypical off-gas treatment components and real-time recycle of the liquid effluents back to the melter feed demonstrated that recycle resulted in significant increases of the incorporation of technetium into the LAW glass product and behavior that was in good agreement with process models for the system. However, the maximum technetium retention achieved in those tests was less than predicted. Tests conducted in the present work demonstrated, for the first time, near-complete retention of technetium in the LAW glass product with recycle. Holdup of technetium in the vacuum evaporator, which concentrates the liquid effluents prior to their recycle to the melter feed, was identified as the primary reason for the previously observed shortfall. The fate of other volatile constituents was also assessed. In particular, in addition to technetium, rhenium, which is frequently used as a non-radioactive surrogate for technetium, was included in these tests. This allows for a direct comparison of the behavior of rhenium and technetium during vitrification and in the down-stream processes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗