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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Monitoring of Actinide Concentrations in Molten LiCl-KCl Salt using Alpha Spectroscopy: Final Progress Report

This project addresses the need for real time monitoring of actinide concentrations for spent fuel reprocessing utilizing alpha spectroscopy. This project specifically focuses on spent fuel reprocessing with the pyroprocessing technique that uses high temperature molten salt. Pyroprocessing uses molten salt such as lithium chloride and potassium chloride to separate out uranium and plutonium from spent fuel. To prevent any possible diversion of high purity plutonium in this process, monitoring of the actinide concentrations within the refinery are essential for safeguards and non-proliferation efforts. Alpha spectroscopy enables the identification and quantification of actinide isotopic concentrations. However, current assay methods involving intrusive sampling and ex-suit measurement require weeks to provide accurate information. A near real-time and in-situ measurement system is required to quickly quantify actinide isotopic concentrations within the salt. Semiconductor radiation detectors provide a compact, high energy resolution option for completing these measurement needs. Silicon carbide (SiC) has been shown to operate effectively at high temperatures and high chemical stability in the molten salt environment. 4H-SiC Schottky diode alpha particle detectors were fabricated and tested to determine their alpha spectroscopy capabilities. The detectors need to operate at the elevated temperatures, survive the molten salt environment, be radiation hard, and have sufficient energy resolution under all these conditions to determine isotopic concentrations. Along with the detector, a method of producing samples of the actinides in the salt is required. To produce a thin film of the actinides on the detector, which is a necessity for alpha spectroscopy, electrodeposition can be used. Electrodeposition uses an electric potential difference to selectively deposit uranium or other actinides from the salt using an anode and cathode. Deposition of the material directly onto the sensor would produce a thin film of material that would enable materials accountability measurements to be taken in real time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

The Future of Nuclear Energy: Electrochemical Reprocessing of Fuel Takes Center Stage

Nuclear power plants use energy-dense fuel and provide dependable baseload energy without generating greenhouse gas emissions. Despite these advantages, the long-term management of used nuclear fuel (UNF) remains a key challenge due to its lifetime (hundreds of thousands of years) and radiotoxicity. The components of UNF that contribute the most to this challenge are the actinide elements. A potential solution to this issue is to separate these radioisotopes from the bulk of the UNF and recycle them as fuel in advanced nuclear reactors. These separations can be achieved using electrochemical reprocessing, which employs electrochemical conversion and electrodeposition in a high-temperature molten salt electrolyte medium to separate the actinides from UNF. In this work, we review the current status, fundamental challenges, and future prospects of electrochemical reprocessing as they relate to UNF recycling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A novel approach to criticality accident detection for a legacy facility

For operations on the Sellafield site involving fissile material a Criticality Accident Alarm System (CAAS) has traditionally been installed unless an omission case can be made. An omission case would examine if, in the absence of all controls, a criticality could reasonably be expected or whether the potential dose to an operator would remain below a maximum acceptable dose. It is a UK legal requirement for adequate emergency arrangements to be in place for hazards on a nuclear licensed site. For facilities where criticality is credible there is a regulatory expectation that a -Criticality Warning System- should be in place. A -Criticality Warning System- is not necessarily a full CAAS and alternative means of criticality detection may be part of an adequate emergency plan. The Sellafield site has a Magnox fuel reprocessing facility which has been operated since the 1960's. Evaporation of the Medium Active (MA) liquid waste from this process is undertaken in an associated building. Following extensive review of the plant and process a number of maloperations which can route fissile material to the MA evaporator were identified. Adequate controls to prevent these scenarios were shown to be in place. In the unlikely event of a failure of controls leading to sustained loss of fissile material a criticality incident may occur and hence an emergency plan was determined to be necessary. The novel emergency arrangements which were developed utilized the existing gamma monitor network located across multiple building floors. These gamma monitors are present for radiological safety. This emergency arrangement would require rapid interpretation of the gamma alarms by an operator and the manual triggering of a criticality evacuation alarm. The gamma monitors were demonstrated to be able to detect the minimum incident of concern (10{sup 15} fissions) as well as being able to alarm during the ramp up in power of a large criticality incident (10{sup 18} fissions). It was shown that a criticality accident could readily be distinguished from a radiological event (e.g. loss of containment). To support this work, criticality excursion analysis was used to examine the likely characteristics of a criticality in the MA evaporator. This novel arrangement was justified due to a number of factors including: the limited remaining operational lifetime of this facility (<5 years), the low frequency of maloperation coupled with a failure of controls and the crucial risk reduction provided by the continued reprocessing of Magnox fuel. Furthermore, the time and cost involved with design and installation of a standard CAAS would be substantially greater than a plan based upon existing instrumentation. These arguments were accepted by the UK regulator the Office for Nuclear Regulation (ONR). (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Assembly and Testing of a Tritium and Iodine Removal System for Use with Advanced Tritium Pretreatment

US regulations will require the removal of iodine and tritium, along with other volatile and semivolatile radionuclides, from the off-gas streams of nuclear fuel reprocessing plants. Tritium pretreatment is a head-end operation that could be incorporated within nuclear fuel reprocessing plants for the upfront removal and trapping of these volatiles species. Quantitative, up-front removal in a single compact system is of significant advantage because otherwise tritium and iodine distributes to other unit operations, each of which have associated off-gas streams also requiring treatment. Oak Ridge National Laboratory is presently developing an advanced tritium pretreatment process based on the oxidation of used nuclear fuel at low temperature using NO 2 /O 2 mixtures, which produces a fine powder. This process can result in the quantitative volatilization of both tritium and iodine. The powder product can be selected to be U 3 O 8 , UO 3 , or a nitrate by adjusting the processing conditions and the process can be readily interfaced to either conventional aqueous-based or advanced pyrochemical–based reprocessing plants. In preparation for a planned kilogram-scale hot cell demonstration of the advanced tritium pretreatment process with used nuclear fuel (UNF), an integrated iodine and tritium capture system was designed and assembled. Three experimental runs of this system were completed successfully. All the samples were submitted for analysis. The results and data analysis will be covered in a subsequent report to be issued by the Off-Gas Sigma Team. Several additional tests should be undertaken to provide additional information on the system’s performance. The objectives for the additional tests and test conditions are presented in this report.

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DOE OSTI2017

Milestone Report - M4FT-17OR030107025 - Design of a tritium and iodine removal system for use with advanced TPOG

US regulations will require the removal of iodine and tritium, along with other volatile and semi-volatile radionuclides, from the off-gas streams of nuclear fuel reprocessing plants. Advanced tritium pretreatment (TPT) is an additional head-end operation that could be incorporated within nuclear fuel reprocessing plants. It utilizes nitrogen dioxide (NOR2R) as an oxidant to convert UOR2R to UR3ROR8R prior to traditional aqueous dissolution. Advanced TPT can result in the quantitative volatilization of both tritium and iodine. Up-front removal of iodine is of significant advantage because otherwise it distributes to several unit operations and the associated off-gas streams. The off-gas streams will then require treatment to comply with US regulations. Advanced TPT is currently under development at Oak Ridge National Laboratory, and a kilogram-scale hot cell demonstration with used nuclear fuel (UNF) is planned for fiscal year (FY) 2018.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Fourier Transform Infrared Spectroscopy and Multivariate Analysis for Online Monitoring of Dibutyl Phosphate Degradation Product in Tributyl Phosphate/n-Dodecane/Nitric Acid Solvent

In liquid–liquid extraction separation processes, accumulation of organic solvent degradation products is detrimental to the process robustness, and frequent solvent analysis is warranted. Our research explores the feasibility of online monitoring of the organic solvents relevant to used nuclear fuel reprocessing. This paper describes the first phase of developing a system for monitoring the tributyl phosphate (TBP)/n-dodecane solvent commonly used to separate used nuclear fuel. In this investigation, the effect of extraction of nitric acid from aqueous solutions of variable concentrations on the quantification of TBP and its major degradation product dibutylphosphoric acid (HDBP) was assessed. Fourier transform infrared (FTIR) spectroscopy was used to discriminate between HDBP and TBP in the nitric acid-containing TBP/n-dodecane solvent. Multivariate analysis of the spectral data facilitated the development of regression models for HDBP and TBP quantification in real time, enabling online implementation of the monitoring system. The predictive regression models were validated using TBP/n-dodecane solvent samples subjected to high-dose external ?-irradiation. The predictive models were translated to flow conditions using a hollow fiber FTIR probe installed in a centrifugal contactor extraction apparatus, demonstrating the applicability of the FTIR technique coupled with multivariate analysis for the online monitoring of the organic solvent degradation products.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2018

Evaluation of the Waste Streams from Management of Reprocessing Off-Gas Streams - 18483

Nuclear fuel reprocessing is being considered in the United States as part of an effort to help meet US energy needs and as a means to improve waste management. During the chemical processing of UNF to recover the valuable components for reuse, a number of volatile radionuclides will be released into the off-gas streams of the reprocessing plant. These volatile radionuclides include H-3, C-14, Kr-85, and I-129. Off-gas abatement systems for the volatile radionuclides will be required to ensure that air emissions from a UNF reprocessing facility can meet or exceed emission requirements set forth by the US Environmental Protection Agency and the US Nuclear Regulatory Commission. An initial engineering evaluation of offgas abatement systems was completed recently. The evaluation addressed the designs of the head-end offgas treatment system, which typically includes both the shear off-gas and the dissolver off-gas and vessel off-gas treatment system. Based on this study, the waste volumes for each of the four primary volatile radionuclides (H-3, C-14, Kr-85, and I-129) were estimated. Included in these estimates were the nonradioactive isotopes and the 'tramp' elements found in air (e.g., Kr and C-12 as CO{sub 2}) and process chemicals (e.g., I and Cl). Because the final waste form for these radionuclides has not been selected, estimated volumes were calculated for various alternative waste forms as appropriate. The resulting waste volume for I was found to be highly dependent on the final waste form selected and increased by a factor of 2 to 4 by tramp halogens that would be released into the off-gas stream from the process chemicals. The volume of tritiated water from the fuel was quite small, but the estimated volume of the final waste form depended on the tritium management approach employed within the reprocessing plant. The Kr waste volume was calculated assuming pure Kr (i.e., assuming complete separation of Xe and Ar). For the case of no Xe separation, the estimated waste volume increased by factor of ∼10. Argon from the air, if not removed, increased the estimated volume by a factor of 425. The estimated volume of C-14 waste was dominated by the CO{sub 2} from the air leaking into the processing cell. (authors)

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DOE OSTI2020

Sol–gel synthesis of iodosodalite precursors and subsequent consolidation with a glass binder made from oxides and sol–gel routes

Radioiodine accumulates in aqueous solutions and off-gas streams during nuclear fuel reprocessing due to its solubility in aqueous solutions and volatility, respectively. Additionally, radioiodine is highly mobile in geological environments and iodine-129 has been found to be one of the largest long-term contributors to chronic dose to humans in scenarios of radionuclide migration from geological repositories. Most of the radioiodine can be captured during fuel reprocessing in off-gas streams using solid sorbents and scrubbing solutions. Once iodine is captured, it must be stored in a durable form for eventual disposal. Iodosodalite has been investigated as a waste form for radioiodine, however these synthesis processes typically result in mixed products and iodine volatilization during consolidation. Therefore, this paper proposes a novel approach to synthesizing iodosodalite utilizing a sol-gel method. This method was chosen to lower processing temperatures and improve product yield. Preliminary experiments conducted to determine the viability of this synthetic method are presented. Additionally, consolidation of sol-gel derived iodosodalite with a glass-binder was explored using three different methods: 1) incorporating the glass-binder during gel preparation using alkoxide precursors; 2) separately preparing the glass binder using a sol-gel method; and 3) separately preparing the glass-binder using a melt-quench technique. Glass-bonded iodosodalite was successfully synthesized using these novel sol-gel approaches.

36 MATERIALS SCIENCE↗
DOE OSTI2017

X-ray Absorption Spectroscopy Investigation of Iodine Capture by Silver-Exchanged Mordenite

Capture of radioactive iodine is a significant consideration during reprocessing of spent nuclear fuel and disposal of legacy wastes. While silver-exchanged mordenite (AgZ) is widely regarded as a benchmark material for assessing iodine adsorption performance, previous research efforts have largely focused on bulk material properties rather than the underpinning molecular interactions that achieve effective iodine capture. As a result, the fundamental understanding necessary to identify and mitigate deactivation pathways for the recycle of AgZ is not available. In this paper, we applied X-ray Absorption Fine Structure (XAFS) spectroscopy to investigate AgZ following activation, adsorption of iodine, regeneration, and recycle, observing no appreciable degradation in performance due to the highly controlled conditions under which the AgZ was maintained. Fits of the extended XAFS (EXAFS) data reveal complete formation of Ag 0 nanoparticles upon treatment with H 2 , and confirm the formation of α-AgI within the mordenite channels in addition to surface γ/β-AgI nanoparticles following iodine exposure. Analysis of the nanoparticle size and fractional composition of α-AgI to γ/β-AgI supports ripening of surface nanoparticles as a function of recycle. Finally, this work provides a foundation for future investigation of AgZ deactivation under conditions relevant to spent nuclear fuel reprocessing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Mechanistic Understanding of Silver Sorbent Aging Processes in Off-Gas Treatment

The effect of aging in various off-gas streams including dry air, humid air, 1% NO/N 2 , 2% NO 2 /air, and a gas mixture consisting of 99.25% humid air (dew point: -15 °C), 0.25% NO/N 2 , and 0.5% NO 2 /air in terms of concentration on reduced silver exchanged mordenite (Ag 0 Z) and silver functionalized silica aerogel (Ag 0 -aerogel) were studied in this project. Utilizing simple and sophisticated experimental information combined with molecular and macroscopic mechanistic modeling, this project has significantly advanced the current understanding of aging for silver bearing adsorbents in nuclear-fuel-reprocessing off-gas streams. In addition, this project has produced models that can be utilized to predict the influence of adsorbent aging on the uptake kinetics and capacity of aged adsorbents for iodine and iodine-containing organic compounds. The results of the study have been published in the open literature, and can be used to help DOE better predict the removal of radioactive gases from nuclear-fuel-reprocessing off-gas streams.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Recent Advances in Radiation-Induced Actinide Redox Chemistry

The actinide series boasts many unique physical and chemical features worthy of both fundamental and applied study. However, the chemical influence of their inherent radiation field is often overlooked, especially as we begin to explore the late actinides in more detail than ever possible before. From the perspective of used nuclear fuel reprocessing, the absorption of ionizing radiation induces the formation of a variety of transient and steady-state excited states, radicals, ions, and molecular degradation products, many of which are highly redox active and can lead to significant changes in a reprocessing solvent system’s physical and chemical properties. For example, radiolysis of the actinides can drive steady-state redox distributions and the formation of non-traditional oxidation states which can complicate their separation and recovery from fission products. This scenario is further exacerbated when complexation is taken into account. Consequently, a molecular-level understanding of radiation effects on the actinides over multiple time, distance, and material domains is essential for supporting innovation in used nuclear fuel reprocessing technologies. Attaining this knowledge necessitates a firm grasp of actinide radiation chemistry to develop predictive, mechanistic, multiscale models to support engineering efforts. Presented here are several recent studies that highlight recent advances in actinide radiation chemistry, in particular, the effect of actinide complexation on ligand reactivity towards radiation-induced transients.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Kinetics for the reaction between the solvated electron and dissolved oxygen in n-dodecane from 2.5 to 40 °C

Temperature-controlled, time-resolved picosecond electron pulse radiolysis was utilized to measure the rate of reaction between the solvated electron (eS–) and dissolved oxygen in n-dodecane solutions from 2.5 to 40 °C for the first time. At 20.0 °C, the reaction rate was determined to be k(eS– + O2) = (4.54 ± 0.21) × 1010 M-1 s-1, with an activation energy of Ea = 14.4 ± 1.3 kJ mol-1. These newly determined kinetic parameters are important for predicting and managing the effects of aerated environments on the degradation of organic solvents used in nuclear fuel reprocessing technologies.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2026

Nanohybrid of Cu2O-Ti3C2Tx as a Silver-Free MXene Sorbent for Iodine Gas Capture from Nuclear Waste

The capture of volatile radioiodine from nuclear fuel reprocessing off-gas streams remains a critical challenge due to the high volatility, long half-life of 129I, and biological uptake of iodide from the environment. Although silver-based sorbents provide strong iodine chemisorption, their high cost and regulatory classification as mixed radioactive-hazardous waste motivate the development of alternative materials. Here, we report a silver-free Cu2O-Ti3C2Tx MXene hybrid for iodine gas capture at 150 °C. Structural and compositional analyses confirm the formation of Cu2O nanoparticles on Ti3C2Tx nanosheets and their subsequent conversion to thermodynamically stable CuI upon static iodine gas exposure, achieving an iodine mass loading of up to 1115 mg/g. These results demonstrate the potential of Cu2O-Ti3C2Tx MXene as a copper-based alternative to silver sorbents for elevated-temperature iodine gas capture.

iodine gas capture↗
DOE OSTI2024

Dodecane Radiolysis Yields by Time-Resolved and Steady-State Methods

Liquid organic molecules are present as solvents, complexing ligands, and additives in both used nuclear fuel reprocessing solvent systems and in their subsequent nuclear waste streams. Under these extreme environments, these organic molecules are constantly exposed to ionizing radiation which promotes their radiolysis, forming a variety of short-lived, highly energetic, excited state and radical species.1-4 Here, we demonstrate new experimental results for the steady-state and time-resolved irradiations of dodecane (C12H26), a long chain, liquid, aliphatic hydrocarbon that is the prototypical solvent used for benchtop studies of aqueous-organic solvent extraction systems. When ionizing radiation interacts with neat dodecane, the energy transfer can result in molecular ionization, to give the dodecane radical cation (C12H26+•) and the solvated electron (eS–), and molecular electronic excitation (C12H26*), which rapidly produces transient carbon-centered radical fragments (CxHy•) and hydrogen atoms (H•).1-4 Studies on the initial yields of the ionization and excitation products were performed using time-resolved picosecond electron pulse radiolysis with the use of molecular probes. Using steady-state cobalt-60 gamma irradiations, the suite of products formed by dodecane radiolysis in aerated and deaerated solutions was determined. Then, using iodine as an alkyl radical scavenger, the loss of molecular iodine with dose was quantified, and by correlating with the molecular hydrogen yields of the system, the initial yields of the various carbon-centered radicals were also determined. Finally, the rates of reactions of the C12H26+• and eS– with ligands proposed for use in spent nuclear fuel reprocessing were studied as a function of temperature from 10 – 40 °C.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗
DOE OSTI2017

Facility Modeling Capability Demonstration Summary Report

A joint effort has been initiated by Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), Savanah River National Laboratory (SRNL), Pacific Northwest National Laboratory (PNNL), sponsored by the National Nuclear Security Administration’s (NNSA’s) office of Proliferation Detection, to develop and validate a flexible framework for simulating effluents and emissions from spent fuel reprocessing facilities. These effluents and emissions can be measured by various on-site and/or off-site means, and then the inverse problem can ideally be solved through modeling and simulation to estimate characteristics of facility operation such as the nuclear material production rate. The flexible framework called Facility Modeling Toolkit focused on the forward modeling of PUREX reprocessing facility operating conditions from fuel storage and chopping to effluent and emission measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Hybrid Ultra‐Microporous Materials for Selective Xenon Adsorption and Separation

Abstract The demand for Xe/Kr separation continues to grow due to the industrial significance of high‐purity Xe gas. Current separation processes rely on energy intensive cryogenic distillation. Therefore, less energy intensive alternatives, such as physisorptive separation, using porous materials, are required. Herein we show that an underexplored class of porous materials called hybrid ultra‐microporous materials (HUMs) affords new benchmark selectivity for Xe separation from Xe/Kr mixtures. The isostructural materials, CROFOUR‐1‐Ni and CROFOUR‐2‐Ni, are coordination networks that have coordinatively saturated metal centers and two distinct types of micropores, one of which is lined by CrO 4 2− (CROFOUR) anions and the other is decorated by the functionalized organic linker. These nets offer unprecedented selectivity towards Xe. Modelling indicates that the selectivity of these nets is tailored by synergy between the pore size and the strong electrostatics afforded by the CrO 4 2− anions.

Mohamed, Mona H.↗
DOE OSTI2017

Water Cleaning System for the Exhaust Pipe of HLW Glass Melter (LFCM) - 17180

Vitrification technology has been generally applied to stabilize High-Level Radioactive Waste (HLW) generated from spent fuel reprocessing. In Japan, Joule Heated Liquid Fed Ceramic Melter (LFCM) system is selected for HLW vitrification and the recent active test has been successfully demonstrated its stable operation. The higher reliability of off-gas treatment system (OGTS) which consists of wet scrubber, absorbers and filters connected to the LFCM is also important to ensure long term system operation. The OGTS removes radioactive gases, mists and dust generated by LFCM and maintains radioactive containments by keeping negative pressure inside. In the past active and inactive tests, it was observed that differential pressure in the exhaust pipe attached to LFCM gradually increased during continuous operation and some deposits were observed inside the off-gas pipe. So the cleaning operation needs to be improved and more frequent in order to avoid the exhaust pipe clogging. Therefore we investigated several cleaning systems and operational methods for the exhaust pipe of LFCM. As a result, we have successfully developed the modified water cleaning system. In this presentation we will introduce our water cleaning system. In addition, we will present the result of inactive mock-up tests for the visualization of this water cleaning system. (authors)

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