Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Fuel Reprocessing”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Structural Investigation of Technetium Dibutylphosphate Species Using X-ray Absorption Fine Structure Spectroscopy

The speciation of Tc after the extraction of Tc(IV) from H 2 O and 1 M HNO 3 by dibutylphosphoric acid (HDBP) in dodecane has been studied by X-ray absorption fine structure (XAFS) spectroscopy. Results show the formation of dimeric species with Tc 2 O 2 and Tc 2 O units, and the formulas [Tc 2 O 2 (DBP·HDBP) 4 ] (1) and [Tc 2 O(NO 3 ) 2 (DBP) 2 (DBP·HDBP) 2 ] (2) were, respectively, proposed for the species extracted from H 2 O and 1 M HNO 3 . The interatomic Tc–Tc distances found in the Tc 2 O 2 and Tc 2 O units [2.55(3) and 3.57(4) Å, respectively] are similar to the ones found in Tc(IV) dinuclear species. It is likely that the speciation of Tc(IV) in dodecane is due to the extraction of a species with a Tc 2 O unit for (2) and to the redissolution of a Tc(IV)-DBP solid for (1). The XAFS results for (1) and (2) were compared to that obtained for the extraction of Tc(IV) with TBP/HDBP/dodecane from 0.5 M HNO 3 , (3) which highlight the formation of Tc mononuclear nitrate species {i.e. [Tc(NO 3 ) 3 (DBP)] or [Tc(NO 3 ) 2 (DBP·HDBP)]}. These results confirm the importance of the preparation and speciation of the Tc(IV) aqueous solutions prior to extraction and how much this influences and drives the final Tc speciation in organic extraction. Here, these studies outline the complexity of Tc separation chemistry and provide insights into the behavior of Tc during the reprocessing of used nuclear fuel.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2026

Influence of Nd(II) and Nd(III) Ions on the Speciation and Kinetic Dynamics of Radiolytic Transients in Molten LiCl-KCl Eutectic Salt Mixtures

In the pyrochemical reprocessing of used nuclear fuel, understanding the effects of ionizing radiation on the chemical behavior of fission products in molten chloride eutectic salt is crucial. Here, this study investigates the impact of radiation on the chemical and kinetic speciation of Nd ions in molten LiCl-KCl eutectic, which are challenging to separate due to their multivalent oxidation states and comproportionation reactions. Electron pulse radiolysis techniques were used to determine transient absorption spectra, chemical kinetics, and activation parameters for the reaction of Nd ions with the primary products of molten LiCl-KCl eutectic radiolysis: the solvated electron (e S - ) and dichlorine radical anion (Cl 2 •- ). Initially, Nd(III) reacted rapidly with e S - , forming Nd(II) with a second-order rate coefficient of k = (4.54 ± 0.07) × 10 10 M -1 s -1 at 673 K. The resulting Nd(II) ions then reacted slowly with Cl 2 •- , regenerating Nd(III) with an estimated rate of k = (1.72 ± 0.04) × 10 10 M -1 s -1 at 673 K. Additionally, the comproportionation reaction of Nd(III) and Nd(s) to form Nd(II) was monitored chemically and electrochemically. This work suggests that the comproportionation reaction of Nd(III) with Nd(s) to form Nd(II) is unstable in molten LiCl-KCl eutectic.

36 - MATERIALS SCIENCE↗
DOE OSTI2018

Molecular Hydrogen Yields from the Alpha Self-Radiolysis of Nitric Acid Solutions Containing Plutonium or Americium

The yield of molecular hydrogen, as a function of nitric acid concentration, from the a-radiolysis of aerated nitric acid and sulfuric acid mixtures containing plutonium or americium has been investigated. Comparison of experimental measurements with predictions of a Monte Carlo radiation track chemistry model shows that, in addition to scavenging of the hydrated electron, its precursor, and the hydrogen atom, the quenching of excited state water is important in controlling the yield of hydrogen. In addition, increases in solution acidity cause a significant change in the track reactions that can be explained as resulting from scavenging of e aq - by H aq + to form H • . While plutonium has been shown to be an effective scavenger of precursors of molecular hydrogen below 0.1 mol dm-3 nitrate, previously reported effects of plutonium on G(H 2 ) a between 1 and 10 mol dm -3 nitric acid were not reproduced. Modelling results suggest that plutonium is unlikely to effectively compete with nitrate ions in scavenging the precursors of molecular hydrogen at higher nitric acid concentrations and this was confirmed by comparing hydrogen yields from plutonium solutions with those from americium. Finally, comparison between the radionuclide, ion accelerator experiments and model predictions leads to the conclusion that the high dose rate of accelerator studies does not significantly affect the measured hydrogen yield. These reactions provide insight into the important processes for liquors common in the reprocessing of spent nuclear fuel and the storage of highly radioactive liquid waste prior to vitrification.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Adsorption of Radioactive Iodine Using Nanocarbon on ETS-10 as Adsorbent

Here, laboratory-synthesized nanocarbon pelletized with titanosilicate (ETS-10) as a support matrix has been investigated for the capture of radioactive iodine present as methyl iodide (CH 3 I) in the off-gas streams produced during aqueous reprocessing of used nuclear fuel. The mass fraction of carbon in the sorbent matrix was 0.10. The effects of residence time and CH 3 I concentration were investigated using a continuous flow column setup to quantify the adsorption and desorption capacities of adsorbent under dynamic conditions from an air stream containing CH 3 I present at concentrations representative of those expected in the off-gas streams. Air with CH 3 I gas as a source in the column resulted in quantifiable CH 3 I adsorption with 0.98 mg/g of adsorption capacity. Laboratory-made nanocarbons had a larger adsorption capacity than those of the other carbons reported in the literature. Additionally, the adsorption capacity of nanocarbon on ETS-10 is compared to that of nanocarbon coated on cordierite in previous studies.

ETS-10↗
DOE OSTI2013

Explorations of TALSPEAK chemistry in extraction chromatography: Comparisons of TTHA with DTPA and HDEHP with HEH[EHP]

An advanced nuclear fuel cycle includes a reprocessing stage for the group separation of the lanthanides from the trivalent actinides. The TALSPEAK (Trivalent Actinide - Lanthanide Separation by Phosphorus reagent Extraction from Aqueous Komplexes) solvent extraction process provides the An3+/Ln3+ separation through the use of an organic lanthanide extractant (bis-2-ethyl(hexyl) phosphoric acid, HDEHP) and an aqueous actinide-selective holdback reagent (diethylenetriamine-N,N,N’,N’’,N’’-pentaacetic acid, DTPA). Organic phase interactions in TALSPEAK have proven sufficiently complex that, even with substantial research effort, a descriptive thermodynamic model has yet to be developed. If one were to consider an extraction chromatographic (EXC) version of TALSPEAK, the presence of concentrated extractant on the resin surface should augment the multifaceted organic phase chemistry. This study examines the previously unexplored impact of each aqueous phase TALSPEAK component (holdback reagent, carboxylic acid buffer, and pH) on the chemistry of the EXC system. The presence of alternative reagents, 2-ethyl(hexyl) phosphonic acid mono-2-ethyl(hexyl) ester acid (HEH[EHP]) and triethylenetetramine-N,N,N’,N’’,N’’’,N’’’-hexaacetic acid (TTHA) are also considered. Results indicate the concentrated extractant appears to enhance solid-phase reactions that most likely involve water, lactate and sodium. The presumed water, lactate and sodium partitioning consumes resin capacity sufficiently that the weaker hold-back reagent under TALSPEAK aqueous conditions (pH ~ 3.6), TTHA, provides preferential Am3+/Ln3+ separations performance. If a TALSPEAK-EXC process were to be developed, a resin with solvent diluted extractant or covalently bound functional groups may be preferential to reduce the complexity of the solid-phase chemistry.

actinide separation↗
DOE OSTI2021

Single-component-at-a-time variation study for glass-ceramic waste forms

Here, a 51-sample composition variation study was performed on glass-ceramic waste forms for a raffinate waste stream from aqueous reprocessing of used nuclear fuel containing high fractions of Mo, alkalis, alkaline earths (AEs), and rare earths (REs). The study was designed with a single-component-at-a-time variation approach off a centroid composition. The components that were varied included Al, B, Ca, Li, Mo, Na, REs, Si, Zr, and Others (containing minor components). Data analysis included crystallization curves, microstructure, and phase compositions. A number of components (i.e., Li 2 O, B 2 O 3 , REO x , MoO 3 , Na 2 O, and ZrO 2 ) significantly impacted the concentration and chemistry of phases, especially the primary phases of oxyapatite [i.e., Ca 2 RE 8 (SiO 4 ) 6 O 2 ] and powellite (i.e., AEMoO 4 ), precipitated in the slow-cool heat-treated waste forms; minor phases included cerianite [i.e., Ce x Zr (1-x) O 2 ], Ba-molybdate [i.e., Ba(Gd 0.67 Mo 0.33 O 3 ], noble metals, pollucite (i.e., CsAlSiO 4 ), and RE-borosilicate (i.e., RE 3 BSi 2 O 10 ).

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

Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors

This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as Pu FP . The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the Pu FP composition. The FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Expanded Analysis of Hot Isostatic Pressed Iodine-Loaded Silver-Exchanged Mordenite

Reduced silver-exchanged mordenite (Ag0Z) is being evaluated as a potential material to control the release of radioactive iodine that is released during the reprocessing of used nuclear fuel into the plant off-gas streams. The purpose of this study was to determine if hot pressing could directly convert this iodine loaded sorbent into a waste form suitable for long-term disposition. The minimal pretreatment required for production of pressed pellets makes hot pressing a technically and economically desirable process. Initial scoping studies utilized hot uniaxial pressing (HUPing) to prepare samples of non-iodine-loaded reduced silver exchanged mordenite (Ag0Z). The resulting samples were very fragile due to the low pressure (~ 28 MPa) used. It was recommended that hot isostatic pressing (HIPing), performed at higher temperatures and pressures, be investigated. HIPing was carried out in two phases, with a third and final phase currently underway. Phase I evaluated the effects of pressure and temperature conditions on the manufacture of a pressed sample. The base material was an engineered form of silver zeolite. Six samples of Ag0Z and two samples of I-Ag0Z were pressed. It was found that HIPing produced a pressed pellet of high density. Analysis of each pressed pellet by scanning electron microscopy-energy dispersive spectrophotometry (SEM-EDS) and X-ray diffraction (XRD) demonstrated that under the conditions used for pressing, the majority of the material transforms into an amorphous structure. The only crystalline phase observed in the pressed Ag0Z material was SiO2. For the samples loaded with iodine (I-Ag0Z) iodine was present as AgI clusters at low temperatures, and transformed into AgIO4 at high temperatures. Surface mapping and EDS demonstrate segregation between silver iodide phases and silicon dioxide phases. Based on the results of the Phase I study, an expanded test matrix was developed to examine the effects of multiple source materials, compositional variations, and an expanded temperature range. Each sample was analyzed with the approach used in Phase I. In all cases, there is nothing in the SEM or XRD analyses that indicates creation of any AgI-containing silicon phase, with the samples being found to be largely amorphous. Phase III of this study has been initiated and is the final phase of scoping tests. It will expand upon the test matrix completed in Phase II and will examine the durability of the pressed pellets through product consistency testing (PCT) studies. Transformation of the component material into a well-characterized iodine-containing mineral phase would be desirable. This would limit the additional experimental testing and modeling required to determine the long-term stability of the pressed pellet, as much of that information has already been learned for several common iodine-containing minerals. However, this is not an absolute requirement, especially if pellets produced by hot isostatic pressing can be demonstrated through initial PCT studies to retain iodine well despite their amorphous composition.

hot isostatic pressing↗
DOE OSTI2018

Design of, and Test Plan for, an Iodine and Tritium Capture System for a NO 2 -Based Tritium Pretreatment Process

The reprocessing of used nuclear fuel would release volatile radionuclides into the off-gas of a processing plant, including 3 H, 14 C, 85 Kr, and 129 I. One potential simplification to the management of the off-gas streams could be achieved through an efficient tritium pretreatment (TPT) step in which the UO 2 fuel is oxidized either by air or by NO2 before dissolution. Oxidation of the UO2 fuel matrix results in the release of tritium contained within. Upfront removal of tritium from the fuel in a pretreatment step, followed by the abatement of tritium, can minimize or eliminate the distribution of tritium throughout the plant, also decreasing or eliminating the need for tritium capture on multiple off-gas streams. The use of NO 2 as the oxidant in an advanced tritium pretreatment (ATPT) allow the oxidation to be performed at lower temperatures and may result in the quantitative release of iodine from the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Testing of an Iodine and Tritium Capture System for an NO₂-Based Tritium Pretreatment Process

The reprocessing of used nuclear fuel would release volatile radionuclides into the off-gas streams of a processing plant, including ³H, ¹⁴C, ⁸⁵Kr, and ¹²⁹I. One potential simplification to the management of the off-gas streams could be achieved through an efficient tritium pretreatment (TPT) step in which the UO₂ fuel is oxidized by either air or NO₂ before dissolution. The oxidation of the UO₂ fuel matrix will result in the release of tritium from the fuel. Upfront release of tritium from the fuel in a pretreatment step with subsequent tritium abatement can minimize or eliminate the distribution of tritium throughout a plant. This can decrease or eliminate the need for tritium capture on multiple off-gas streams and prevent distribution to the aqueous inventory. The use of NO₂ as the oxidant in an advanced tritium pretreatment (ATPT) allows for the oxidation to be performed at lower temperatures and may also result in the quantitative release of iodine from the fuel. The ATPT off-gas will contain up to 75 vol% NO₂ gas. Spencer et al. identified the most promising iodine sorbent as silver nitrate-impregnated alumina (AgA) and the most promising tritium sorbents as 3Å molecular sieve (3Å MS) and silica gel for radionuclide capture from ATPT off-gas streams. An initial series of tests were conducted in fiscal years 2017–2018 with the intent of demonstrating the recovery of iodine and tritium using these sorbents from a recirculating gas stream that simulated the NO₂ recycle loop envisioned for the ATPT system. These initial tests showed low tritium recoveries and significant variations in the iodine recovery on the AgA. Ultimately it was concluded, following disassembly of the system, that corrosion within the system was partially the cause of the poor recovery of both species. All components of that system were disposed of at the end of the analysis. One of the conclusions drawn from the initial tests pointed to the reconsideration of the materials of construction. Based on that recommendation a new test bed for the advanced TPT off-gas system, that addresses the issues identified in the fiscal years 2017 and 2018 scoping tests, was designed and constructed. The test bed used high nickel alloys wherever possible to address the previously observed corrosion issues, a redesigned surge tank to permit ease of purging, an improved feed injection system and a more modular system design that allowed testing of incremental increases in system complexity to allow the isolation of any element that results in changes to recovery of target species. Accompanying that design, a multifaceted test plan was completed. The overarching goals of this test series were two-fold. The first goal was to demonstrate the ability to recover both the iodine and tritium that would be released during TPT in a manner that the releases from the fuel can be quantified. The second goal was to demonstrate that the combined recovery system could operate in any of the three proposed TPT modes (once-through air or O₂, recirculating air or O₂, or recirculating NO x ). A total of 22 tests were completed. Ten tests examined tritium behavior under varying operating parameters, 8 tests examined iodine behavior under varying operating parameters, and 4 tests examined both tritium and iodine behavior. Silica gel was used for the tritium sorbent, and AgA or Ag-mordenite (AgZ) was used as the iodine sorbent. The following conclusions were drawn: Full recovery is achieved for both sorbates when no sorbent is in place indicating no accumulation within the test system. In even the simplest of the configurations (i.e., no sorbents in the columns), iodine is more difficult to transport and recover from the system than tritium. Analysis of the AgA iodine sorbent beds from the integrated tritium and iodine tests indicate quantitative iodine recovery can be achieved (within anticipated experimental errors). This system may provide a means for the quantification of the iodine released by the ATPT process within the experimental uncertainties. This would support 0.33 to 1.00 kg demonstrations of the ATPT process with actual used nuclear fuel. Iodine penetrated further into AgZ than into AgA when operated in gas recirculation mode. Tritium adsorbs to and cannot be fully recovered from the 10-cm column of silica gel at regeneration temperatures up to 140 °C.; Tritium adsorbs to and cannot be recovered from the 10-cm column of AgA at desorption temperatures up to 200 °C. Co-adsorbed tritium was recovered to a greater extent from AgZ than from AgA, but recovery was not complete. No visible corrosion was observed in the test system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Effect of NO x and Water Variations on Iodine Loading of AgZ

Reprocessing of used nuclear fuel will result in the release of several volatile radionuclides that need to be removed from facility off-gas streams before their release to the environment. Iodine, one of these radionuclides, is expected to be primarily released in the dissolver off-gas stream as iodine (I 2 ) or methyl iodide (CH 3 I). Sorbents that target I 2 and CH 3 I removal from the DOG stream will need to adsorb iodine under elevated temperatures and in the presence of water vapor and nitrogen oxides (NO x ). Previous studies examining the adsorption of CH 3 I and I 2 in the presence of NO x gases by silver-exchanged mordenite (AgZ), a zeolite mineral considered for use in this application, have resulted in slightly inconsistent conclusions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Comparison of Extended and Accelerated VOG Tests

The reprocessing of used nuclear fuel would release volatile radionuclides into the off-gas streams of a processing plant, including ¹²⁹I. The dissolver off-gas and more dilute vessel off-gas streams (VOG) must both be targeted when mitigating ¹²⁹I environmental release because each contains some amount of iodine. Iodine-129 will likely be found as elemental iodine (I₂) and methyl iodide (CH₃I) in these off-gas streams. Reduced silver-exchanged mordenite (AgZ or Ag⁰Z) has been investigated as a potential sorbent for iodine abatement and is studied here under prototypic VOG conditions. Because of the relatively low iodine abundance and high flow rates of the VOG, total iodine concentrations are expected to be in the parts per billion range. Thus, VOG experiments need to run for extended durations at low concentrations to reach sorbent saturation. Because the sorbent will be exposed to oxidizing gas streams for extended periods of time, aging effects and sorbent degradation need to be considered when designing a sorbent-based abatement system. Three sets of experiments were performed with the aim of determining how the adsorption of iodine by AgZ is affected by differing test durations and gas compositions. The first tested the capacity of sorbent aged for 8 months under a humid air stream. The second tested differences in sorbent capacity and mass transfer zone (MTZ) length during high-concentration (1200 ppb v ) CH₃I loading in a humid nitrogen gas (N₂) stream and a humid air gas stream over 28 days. The third tested sorbent capacity and MTZ length during low concentration (< 200 ppb v ) CH₃I and I₂ loading in a humid air stream over 9 months. The results of these tests suggest that AgZ capacity drops by ~50% after 1 month of aging, by ~60% after 2 months of aging, and then does not continue to decrease significantly at up to 8 months of aging. One-month aging tests that compared N₂ and air as the gas stream diluent resulted in similar maximum loading capacities and overall loading curves, indicating that the effects of aging cannot be mitigated by avoiding air as the balance gas. The 9-month extended VOG tests did not result in clear sorbent saturation at the inlet, but it can be inferred using an assumption of a maximum capacity of 45 mg I/g AgZ. Applying this assumption, then the MTZ of CH₃I is 12.8 cm, and the MTZ of I₂ is 12.4 cm. These results are similar to previous tests. Comparisons to tests completed at Idaho National Laboratory suggest that the CH₃I MTZ may be dependent on concentration. Future work should re-evaluate the design of the VOG iodine capture system with this updated data and should seek to understand fundamental characteristics of CH₃I adsorption by AgZ, such as the effect of concentration, gas velocity, and gas composition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Actinide N-Donor Thermodynamics: Expanding the f-element Covalency Dialogue. Final report

Resolving the chemistry and physics of f-electrons is one of the grand challenges of science for energy technology. A central component of this challenge is describing covalency in actinide-ligand interactions. Crystallography, x-ray spectroscopy and computational techniques have provided reasonable, and in some cases remarkable, support for covalency in actinide-ligand interactions. Courtesy the quality of the results obtained, a significant portion of the covalency dialogue has been forwarded through solid-state and computational studies of the more stable actinides (Th, U, Pu, Np). However; recent reports have questioned whether covalency in actinide interactions increases or decreases across the series. Examination of less-stable trans-actinides (Am, Cm, Bk, Cf and Es) by the aforementioned approaches can be rapidly limited due to material availability, radiological hazards and experimental data to validate computational models. Furthermore, limited thermodynamic data exists to confirm covalency in actinide solution phase interactions that are highly relevant to the remediation and reprocessing of used nuclear fuel. The best-defined interactions of actinides with soft donors in aqueous solutions involve (poly)aminopolycarboxylate (APC) ligands. The chelate effect and binding affinity with the acetic acid APCs subgroups encourages amine interactions with the actinide metal center. In the absence of these factors, amine interactions with actinides are too weak to overcome the protective hydration shell of the dissolved ion. The ability for APCs to force actinide interactions with soft nitrogen donors (as defined by Pearson’s Hard Soft Acid Base theory) encourages the application of these ligands in a variety of processes for actinide recovery from nearly chemically identical lanthanides. The ability to functionalize the amine center of the ligand in a variety of capacities (adding additional amine groups, exchanging the conventional acetate group for an acetate group, etcetera), allows for the APC ligand to serve as a thermodynamic probe for actinide-nitrogen interactions. Perhaps the single most significant breakthrough during the previous funding cycle was the observation that covalency for the transplutonium part of the actinide series can be increasingly influenced with energy degeneracy driven covalency as the actinides become heavier. This was observed most predominantly with dipicolinic acid, but extensions of f-orbital degeneracy were found to affect aliphatic aminopolycarboxylate-actinide complexes through einsteinium.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Analysis of Organoiodide Adsorption Mechanisms

If the United States were to engage in the reprocessing of used nuclear fuel, radioactive iodine must be removed from multiple plant off-gas streams to comply with governing regulations. One of these streams is the vessel off-gas (VOG), which arises from the separations process and is expected to contain iodine in primarily organic iodine forms and at parts-per-billion concentrations. The relative lack of knowledge surrounding organic iodine removal from the VOG prompted the US Department of Energy’s Office of Nuclear Energy to initiate experimental efforts targeted at understanding organic iodide removal from prototypic VOG streams. At Oak Ridge National Laboratory (ORNL), this effort has focused on developing a comprehensive understanding of iodine removal from VOG streams by using silver-based sorbents. An important aspect of this testing is developing an understanding of how the sorption of methyl iodide (CH 3 I), the most studied organic iodide species to date, compares with the sorption of other volatile organic iodide species potentially present in the VOG. The work presented here reflects initial testing that will continue to be developed, and final results will be incorporated into an end-of-year report that details a multiyear testing campaign designed to understand iodine mitigation from VOG streams. The goal of these experiments was to determine whether similar reaction pathways govern both CH 3 I and iodobutane (C 4 H 9 I) sorption onto silver mordenite (AgZ), a common silver-based iodine sorbent. More specifically, the authors hypothesized that the sorption of C 4 H 9 I by AgZ will result in the formation of butanol (C 4 H9OH). Effluent monitoring of CH 3 I sorption testing has confirmed methanol production, but analogous monitoring of effluents from C 4 H 9 I sorption studies has not been performed. A series of tests was completed to test this hypothesis with the aim of detecting C 4 H 9 OH downstream of the AgZ bed. Test conditions varied the bed depth, gas stream humidity, and bed temperature post sorption. The effluent gas stream downstream of the AgZ bed was sampled by using gas-tight syringes, and these samples were analyzed by a gas chromatograph coupled to a mass spectrometer. Thin bed C 4 H 9 I tests conducted at -65 and 0°C dew points did not result in C 4 H 9 OH detection in the effluent. This suggests that the sorption mechanism and subsequent reactions could be different from those observed for CH 3 I. The sorption testing continues; additional results will be included in final end-of-year report and will increase the fundamental understanding of organic iodide sorption by AgZ from VOG streams.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Solvent Recovery and Management at the Savannah River Site H-Canyon Facility

NIOWAVE, Inc. is a domestic supplier of medical and industrial isotopes from uranium and radium. The Savannah River National laboratory (SRNL) is currently providing support to NIOWAVE, which plans to deploy a superconducting electron linear accelerator (LINAC) to fission uranium for Mo-99 production without the need for a nuclear reactor or HEU. The uranium from the Mo-99 production targets will be purified using a modified PUREX (Plutonium Uranium Reduction Extraction) solvent extraction process to recover the uranium in the product stream. The uranium will then be precipitated as an oxalate which is calcined to U 3 O 8 to fabricate pellets for new Mo-99 targets. In previous support provided to NIOWAVE, the SRNL demonstrated a solvent washing process to remove degradation products from the tributyl phosphate (TBP) solvent used in the modified PUREX process under development for uranium recovery. To supplement this technology demonstration, NIOWAVE requested the SRNL to provide summary information on the solvent recovery and management activities which are used at the Savannah River Site (SRS) H-Canyon facility. An existing reference document for the reprocessing of irradiated HEU fuels at the SRS was used as the primary reference for the solvent management activities; although, other reference documents were used to provide supplementary information. The information provided includes a brief summary of the solvent degradation issues which have been observed in the H-Canyon solvent extraction cycles and resulting process safety concerns. The solvent recovery processes for the three cycles of solvent extraction used in the H-Canyon were subsequently described including the process equipment which consists of the continuous and batch solvent washers, pumps, and tanks. A final section is provided on the monitoring and analysis of solvent quality based on the previous work performed at the SRNL for NIOWAVE and past research and development activities performed to support the solvent extraction processes in both the SRS F-Canyon and H-Canyon facilities.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Iodine Capture Studies of Silver Mordenite and Novel Alternative Metal Sorbents

Radioiodine is one of the radionuclides of concern when considering reprocessing of used nuclear fuel (UNF). It is expected to be released primarily in the dissolver off-gas (DOG) stream. Sorbents that target I 2 removal from a UNF processing off gas must perform efficiently under elevated temperatures in the presence of water vapor and nitrogen oxide gasses (NOx). Many studies have been conducted that examine the adsorption of iodine species in the presence of NOx gases through reduced silver-exchanged mordenite (Ag 0 Z), a zeolite mineral. This study characterizes the effects of iodine, water vapor, and nitrogen dioxide on the adsorption of I 2 by Ag 0 Z at lower temperatures suitable for comparison with novel Cu- and Bi- bearing sorbents which offer a non-RCRA alternative. When exposed to ‘harsh’ conditions such as I 2 , H 2 O, and NO 2 vapors carried by air, Cu 2 S-polyacrylonitrile (PAN) is more efficient than Ag 0 Z and all other novel sorbents, providing with 11% I capture efficiency. However, under ‘ideal’ conditions in which I 2 is carried by dry air), Ag 0 Z is more efficient than Cu 2 S-PAN and other novel sorbents, providing 74% I capture efficiency. To investigate this result and the overall performance of all sorbents tested, several characterization techniques were employed, including thermal gravimetric analysis, scanning electron microscopy – electron dispersion X-Ray spectroscopy, powder X-ray diffraction, and X-ray photoelectron spectroscopy. These techniques are discussed within this report. This document also includes a design and test plan update from Idaho National Laboratory as the next step in testing the high performing sorbents from thin-bed tests carried out by Oak Ridge National Laboratory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Aluminophosphate Waste Forms for Immobilizing Cations from Electrochemical Salt Wastes

This report provides experimental details and results of evaluating aluminophosphate waste forms for treating and immobilizing the salt cations from salt wastes generated during electrochemical reprocessing of used nuclear fuel. In the waste form process for these materials, chloride salt streams are reacted with NH 4 H 2 PO 4 , the chlorine is removed from the salts and driven off as NH 4 Cl (a solid condensate that can be captured), and then the product can be vitrified in conjunction with glass-forming chemicals (e.g., Fe 2 O 3 , Al 2 O 3 ) to create a high-durability waste form. This study was initiated with some literature review on aluminophosphates containing high alkali oxide content and some of this information is summarized in this report. Following literature review, three new samples were synthesized where two contained Fe 2 O 3 +Al 2 O 3 (i.e., samples G3 and G5 ) and one was only Al 2 O 3 (Fe 2 O 3 -free) (i.e., sample G6 ). In addition to these samples, G1 was also made, which is the baseline reference waste form referred to as DPF5-336 (made without Al 2 O 3 ). Samples G1, G3, G5, and G6 had phases of Li 3 Fe 2 (PO 4 ) 3 (likely), monazite (below XRD detection limits), AlPO 4 , and AlPO 4 , respectively. Characterizations on these materials included optical images, scanning electron microscopy, energy dispersive X-ray spectroscopy, and X-ray diffraction. These samples were shipped to Argonne National Laboratory for chemical durability testing. Depending on how the samples perform in these tests, an additional phase of aluminophosphate formulations could be designed and tested. This report completes the milestone M4FT-23PN030104041 with details provided in Appendix B.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗