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At least 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Investigating the Durability of Iodine Waste Forms in Dilute Conditions

To prevent the release of radioiodine during the reprocessing of used nuclear fuel or in the management of other wastes, many technologies have been developed for iodine capture. The capture is only part of the challenge as a durable waste form is required to ensure safe disposal of the radioiodine. This work presents the first durability studies in dilute conditions of two AgI-containing waste forms: hot-isostatically pressed silver mordenite (AgZ) and spark plasma sintered silver-functionalized silica aerogel (SFA) iodine waste forms (IWF). Using the single-pass flow-through (SPFT) test method, the dissolution rates respective to Si, Al, Ag and I were measured for variants of the IWFs. By combining solution and solid analysis information on the corrosion mechanism neutral-to-alkaline conditions was elucidated. The AgZ samples were observed to have corrosion preferentially occur at secondary phases with higher Al and alkali content. These phases contained a lower proportion of I compared with the matrix. The SFA samples experienced a higher extent of corrosion at Si-rich particles, but an increased addition of Si to the waste led to an improvement in corrosion resistance. The dissolution rates for the IWF types are of similar magnitude to other Si-based waste form materials measured using SPFT.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

Development of the Actinide-Lanthanide Separation (ALSEP) Process

Separating the minor actinide elements (Am and Cm) from acidic high-level raffinates arising from the reprocessing of irradiated nuclear fuel is an important step in closing the nuclear fuel cycle. Most proposed approaches to this problem involve two solvent extraction steps: 1) co-extraction of the trivalent lanthanides and actinides, followed by 2) separation of the actinides from the lanthanides. The objective of our work is to develop a single solvent-extraction process for isolating the minor actinide elements. We report here a solvent containing N,N,N',N'-tetra(2 ethylhexyl)diglycolamide (T2EHDGA) combined with 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) that can be used to separate the minor actinides in a single solvent-extraction process. T2EHDGA serves to co-extract the trivalent actinide and lanthanide ions from nitric acid solution. Switching the aqueous phase chemistry to a citrate buffered solution of N-(2-hydroxyethyl)ethylenediamine-N,N',N'-triacetic acid at pH 2.5 to 4 results in selective transfer of the actinides to the aqueous phase, thus affecting separation of the actinides from the lanthanides. Separation factors between the lanthanides and actinides are approximately 20 in the pH range of 3 to 4, and the distribution ratios are not highly dependent on the pH in this system.

actinide separation↗
DOE OSTI2018

Application of a Triple Bubbler Sensor for Determining the Density, Surface Tension, and Depth in Molten Salts

Electrochemical (i.e. pyroprocessing) technology is being widely studied throughout the world as a potential alternative to aqueous reprocessing of spent nuclear fuel (SNF). Central to this process is the electrorefiner (ER), in which the useful uranium in the SNF is electrochemically transported through a molten salt electrolyte to a cathode for later recovery. As part of this process, uranium, plutonium, and other actinides accumulate in the ER salt over time. Material accountancy and safeguards of these special nuclear material buildups in near real-time is a significant challenge due to the elevated processing temperatures, remote operation within a hotcell, and high radiation fields. Analytical techniques outside of the hotcell can be used to provide the SNM concentrations of salt samples from the ER. However, determining the actual mass of these materials in the ER presents a significant challenge because the density and volume within the vessel is largely unknown. The Idaho National Laboratory (INL) has developed a triple bubbler sensor capable of determining the density, surface tension, and depth of a fluid in near real-time. Previous experiments in aqueous media has shown that the sensor is capable of determining the density, surface tension and depth with accuracies on the order of 0.2%. The goal of this current work is to test the sensor in several molten salt media to calibrate and validate the triple bubbler. Experiments have been performed in LiCl-KCl eutectic salt. Calibration factors were 1.0015 and 1.26 for density and surface tension, respectively. With the above calibration factors applied, the depth of the salt in the test vessel was determined to within 0.4%. These results indicate that the triple bubbler sensor has the potential to accurately (less than 1% uncertainty) determine the density, surface, tension, and depth of the ER salt in near real-time. The triple bubbler sensor designed at INL can significantly enhance the material accountancy, safeguards, and process monitoring of the electrochemical processing of SNF.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2020

Radiological Degradation of Tetraoctyl Diglycolamide in n-Alkyl Solvents: Influence of Solvent Ionization Potential

Reprocessing of used nuclear fuel reduces the requirements for long-term storage repositories. Cost-efficient processes depend on knowledge of radiation effects, but gaps occur in understanding how radiation results in the degradation of fuel separation molecules. This project investigates the influence of different n-alkyl solvents on the radiolytic fragmentation behavior of N,N,N',N'-Tetraoctyl-3-oxapentane-1,5-diamide (TODGA), a promising ligand for extraction of the lanthanides and minor actinides from dissolved nuclear fuel. Findings from this work suggest a linkage between the ionization potential of the solvent and the degradation products observed in the resulting mass spectrum.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Crystalline Aluminosilicates for Radionuclide Remediation

Crystalline aluminosilicates have a wide range of applications such as media for the capture and/or immobilization of radionuclides. Several different types of metal-exchanged zeolites have been successfully demonstrated as effective sorbents for gaseous, aqueous, or solid-state capture and immobilization of radionuclides including iodine [e.g., I2(g), CH3I(g)], xenon, krypton, radium (e.g., 226Ra and 228Ra), selenium (e.g., 79Se6+), technetium (e.g., 99TcO4-), strontium (e.g., 90Sr2+), cesium (e.g., 135Cs+, 137Cs+), mixed alkali halide streams (i.e., iodide and chloride salts), and actinides (e.g., U6+, Th4+). These zeolites include sodalite (SOD), mordenite (MOR or Z-type), faujasite (FAU), Linde type-A (LTA), cancrinite (CAN), gismondine (GIS), and heulandite (or clinoptilolite) (HEU). The structural building blocks of these seven materials vary from cages (e.g., a-cages, ß-cages, e-cages) to channels linking various cage types and these are mostly comprised of aluminosilicates. This chapter will delve into the crystallographic differences between these framework materials and provide a brief overview of literature data where these seven crystalline aluminosilicates have been utilized for the capture and/or immobilization of radionuclides. Metal-exchanged zeolites (e.g., MOR, FAU, LTA) have often been the source of study for gaseous capture of iodine where the more promising compositions include Ag-exchanged sorbents; these types of materials can be made in a variety of robust forms including berl saddles and extrudates. Sodalite can be used as an immobilization strategy for halide waste streams (e.g., Cl, I), including those from electrochemical reprocessing of used nuclear fuel, as well as TcO4–. Cancrinite has been demonstrated to enhance sorption capacity of sediments for several problematic ionic radionuclides including Cs+, SeO42–, and TcO4–. Actinides such as U6+ and Th4+ have been successfully captured from solutions using MOR, FAU, LTA, and HEU-type aluminosilicates. Finally, mixtures of aluminosilicates have been included in a set of zeolite materials that show promise as sorbents for 226Ra and 228Ra removal from mine waters.

zeolites, aluminosilicates, radionuclide remediati↗
DOE OSTI2013

A micro hot test of the Chalmers-GANEX extraction system on used nuclear fuel

In the present study, a 'micro hot test' has been performed using the Chalmers-GANEX (Group Actinide Extraction) system for partitioning of used nuclear fuel. The test included a pre-extraction step using N,N-di-2- ethylhexyl-butyramide (DEHBA) in n-octanol to remove the bulk part of the uranium. This pre-extraction was followed by a group extraction of actinides using the mixture of TBP and CyMe{sub 4}-BTBP in cyclohexanone as suggested in the Chalmers-GANEX process, and a three stage stripping of the extracted actinides. Distribution ratios for the extractions and stripping were determined based on a combination of γ- and α-spectrometry, as well as ICP-MS measurements. Successful extraction of uranium, plutonium and the minor actinides neptunium, americium and curium was achieved. However, measurements also indicated that co-extraction of europium occurs to some extent during the separation. These results were expected based on previous experiments using trace concentrations of actinides and lanthanides. Since this test was only performed in one stage with respect to the group actinide extraction, it is expected that multi stage tests will give even better results. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Current status of research and development program for characterizing fuel debris at Fukushima Daiichi NPS by JAEA

Japan Government and TEPCO submitted a research road map for decommissioning Fukushima Daiichi Nuclear Power Plant. Two projects about debris are in progress: 'Assessment of simulated fuel debris characteristics' and 'Development of technologies for the processing of fuel debris'. The major results concerning the first project are the following 4 points. First, it was suggested that typical phase of oxide of fuel debris is (U,Zr)O{sub 2} and that of metal is Fe{sub 2}(Zr,U) by thermodynamic calculation. Secondly, important properties of fuel debris for developing defueling tools were identified as shape, size, density, hardness, elastic modulus, fracture toughness, thermal conductivity, specific heat (heat capacity), and melting point. Thirdly, the influence of seawater salt and B{sub 4}C/SUS to characteristics of debris was found, such as deposition of magnesium oxide crystal on the surface of fuel debris. The Influence of Pu to thermal properties of fuel debris was found, such as the increase of melting point. Concerning the second project, the major results are the following. First, a draft of the whole image of scenarios was developed. Secondly, the alkaline resolution method using Na{sub 2}O{sub 2} is most likely to be applied as a part of analysis technologies. Thirdly, it was shown that a part of fuel debris rich in U might be soluble in nitric acid. Fourthly, it was shown that all pyrochemical processes examined have potential to be applied for treating fuel debris. The results of the projects will contribute to the decommissioning program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

The Effect of Kr-85 Decay to Rb-85 on Waste Forms - 17105

Krypton-85 (Kr-85) is a volatile radionuclide released during the reprocessing of used nuclear fuel (UNF). Prior to discharge of process off-gas streams to the environment, removal of Kr-85 is required to meet emission regulations. Once removed from the off-gas streams, Kr (containing stable Kr in addition to Kr-85) must be placed in storage until the Kr-85 has decayed to levels such that its potential release would meet regulatory guidelines. The current method of storage is high-pressure metal canisters. Historically, concerns have been expressed that the decay product of Kr-85, Rubidium-85 (Rb-85), could cause corrosion and weakening in the canisters that would ultimately lead to premature failure and release of Kr-85. Alternate methods and materials, such as matrix isolation in a vapor deposited metal and immobilization in zeolites, have been investigated as storage approaches for Kr. However, even these storage methods may become compromised due to the chemical effects of Rb-85 arising from the decay of Kr-85. A review was conducted of the available literature to examine topics relevant to the effects of Rb on the long-term storage of Kr-85, examine the source terms in the context of this concern, and suggest a path forward to resolve outstanding knowledge gaps. In the course of this analysis, several knowledge gaps were identified. Examples include: o The use of solid sorbents to potentially permit more Kr-85-bearing gas to be loaded into canisters, either at the pressures used in the current method or at significantly lower pressures. o A cost-to-benefit analysis is needed to ascertain the practicality of manufacturing waste forms for Kr-85 storage with solid sorbents. o The effects of radiation and chemical transmutation in solid sorbents or matrices for this application are virtually unknown. o Fundamental data are needed on corrosion rates and mechanisms as functions of Rb concentration, storage temperature, etc., for various storage approaches (e.g., as compressed gas or encapsulated in a getter material) and storage container materials. The present analysis focused on the available information regarding storage of Kr-85 in high-pressure containers. It was judged unlikely that such containers would be compromised by corrosion from the Rb decay product, especially if the Kr recovery operations result in a gas composition with very low oxygen content. However, this conclusion should be revisited based on preliminary data gathered from analysis of legacy Kr waste forms. These legacy waste forms are a result of experiments conducted at Idaho National Laboratory over 30 y ago and have been stored for more than three Kr-85 half-lives. Evaluation of these forms indicates significant corrosion, the cause of which has not been conclusively determined. (authors)

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

Progress in Defining the UK Highly Active Storage Tanks POCO Strategy - 16134

In the UK Highly Active (HA) waste is stored in stainless steel tanks prior to encapsulation in glass in one of three operating vitrification plants. Significant progress has been made in recent years towards the reduction in the volume of HA Liquor (HAL) stored in these tanks and a number of the tanks are now operating at heel levels. As a result, a strategy is being developed to carry out Post Operational Clean Out (POCO) of some tanks while reprocessing operations continue, thus enabling accelerated and more cost effective clean-up of these HA facilities. This paper, and subsequent conference presentation, will highlight some of the key aspects of this strategy, including a description of the challenges and progress made so far to help facilitate the clean-up mission without having a detrimental effect on ongoing reprocessing operations. (authors)

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

Re-evaluation of the Gamma Radiolysis Radiation Chemistry of Neptunium

The reprocessing of spent nuclear fuel by solvent extraction techniques starts with dissolving fuel in concentrated nitric acid (HNO{sub 3}). The resulting radioactive liquor is subjected to multiple stages of solvent extraction to partition desired metal ions (e.g. uranium, plutonium, and neptunium) into an organic phase using specialized extractants. Under process conditions the solvent system is continuously exposed to a high intensity radiation field, which induces radiolytic degradation of both the diluent and extractants. This degradation occurs through the formation of radicals, as well as degradation products. Moreover, both radicals and degradation products can undergo redox chemistry with metal ions within the solvent system. This can be particularly problematic for metals such as neptunium which depend on the maintenance of a particular valence state amenable to complexation by extractants such as tributylphosphate in the PUREX (Plutonium Uranium Redox Extraction) process. Neptunium (Np) is well extracted when it is Np(IV) and Np(VI) but almost unextractable when Np(V). A recently study showed that the radiolysis of Np (VI) in acidic (4 M HNO{sub 3}) solution underwent reduction to Np(V). This reduction only occurred after sufficient production of nitrous acid (HNO{sub 2}) which countered the action of radiolytically-produced oxidizing free radicals (nitrate radical, NO{sub 3}{sup .} and hydroxyl radical, {sup .}OH). This radiation-induced chemistry was successfully modeling using a simplified kinetic model, that included a reaction set for water radiolysis, plus published rate constants for the reactions of free radicals with Np(VI) and Np(V). The most sensitive reaction rate constants were identified in this model, and optimization of the five most sensitive was performed to get excellent model agreement with the experimental data. No Np(IV) was observed or predicted by the model under these conditions. The present research re-evaluates this radiation chemistry of neptunium in aerated aqueous solutions of HNO; by using a far more comprehensive FACSIMILE code. We have also extended this modeling to study the radiolytically-induced Np redox chemistry occurring under multiple other nitric acid acidities, to fully characterize the importance of the oxidizing radical and reducing nitrous acid reactions under a range of conditions relevant to large-scale process conditions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Mox powder calculation improvement: OECD Benchmark proposal and cristal - Rib results - 14228

In the framework of the WPNCS Expert Group on Uncertainty Analysis for Criticality Safety Assessment, we propose a blind numerical benchmark. Since the WPNCS Expert Group on Experimental Needs had highlighted the poor database of representative experiments of low-moderated MOX powders, this blind benchmark is focused on MOX wet powders Three PuO{sub 2} contents are proposed: 100%, 30% and 12.5%. Two Pu isotopic vectors are considered: a conservative one (0%, 71%, 17%, 11%, 1% for {sup 238}Pu up to {sup 242}Pu) and a realistic one obtained from reprocessed LWR 30 GWd/T fuels. For these 6 cases the powder moisture rate is 3%. The MOX fissile medium is a sphere surrounded by 20 cm water reflector. The required results are the calculated k{sub eff} and the associated prior uncertainty. Using benchmarks experiments, contributors should provide the projected-k{sub eff} and the corresponding posterior uncertainty. CRISTAL-V2 results based on JEFF3.1.1 nuclear data are presented. As an example, in the 30% PuO{sub 2} challenging case, the calculation uncertainty is decreased from 1160 pcm (1 σ) down to 450 pcm using EOLE and MASURCA critical Pu experiments. This uncertainty depends slightly on 'conservative' or 'realistic' Pu because the representativity factor does not vary significantly with the Pu vector. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Acoustic Monitoring of Pyroprocessing for Safeguards

As pyroprocessing continues to be an attractive option for the reprocessing of spent nuclear fuel worldwide, safeguards technologies are needed to address the monitoring capabilities that can help state level authorities, or the International Atomic Energy Agency (IAEA) maintain continuity of knowledge of the plant operations. Idaho National Laboratory (INL) is studying the possibility of using acoustic monitoring as a means to monitor a pyroprocessing facility for safeguards purposes. This paper discusses the experimental design and some preliminary results of tests conducted at the Fuel Conditioning Facility, a pyrochemical capable facility, at INL.

98 - NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL↗
DOE OSTI2013

Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410

Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for use as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions were agreed. These include an assumed reprocessing capacity of 800 tonnes per year, the requirement to remove as waste forms the volatile fission products carbon-14, iodine-129, krypton-85, tritium and ruthenium-106, the restriction of discharge of any water from the facility unless it meets US Environmental Protection Agency drinking water standards, no intentional blending of wastes to lower their classification, and the requirement for the recovered uranium to be sufficiently free from fission products and neutron-absorbing species to allow it to be re-enriched and recycled as nuclear fuel. The results from this work showed that over 99.9% of the radioactivity in the UNF can be concentrated via reprocessing into a fission-product-containing vitrified product, bottles of compressed krypton storage and a cement grout containing the tritium, that together have a volume of only about one eighth the volume of the original UNF. The other waste forms have larger volumes than the original UNF but contain only the remaining 0.1% of the radioactivity. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Why reconsider the thorium fuel cycle?

In this paper we have endeavored to present the available technical information on the potential use of Th in nuclear fuel cycle (FC) applications as compared to U without subjective evaluations. Where helpful, we have compared the technical attributes of Th-232 as a fertile isotope and U-233 as a fissile isotope with other similar isotopes (i.e., U-238, and U-235 and Pu-239, respectively). In addition, we have summarized (a) experience gained to-date with fabricating and reprocessing of Th-232/U-233 fuels, (b) factors concerning Th fuel irradiation in both test reactors and power reactors, and (c) differences in the backend of the FC with emphasis on repository risks. As might be expected, many technical aspects of Th vs. U have not changed since the sixties. However, there are some factors elaborated in this paper that have changed. Changes potentially encouraging Th use are: (a) the ability to recover large amounts of Th as a byproduct with small attendant costs and environmental impacts, (b) the potential to produce fewer minor actinides (MA) and less Pu during power production, and (c) increased concerns about proliferation which might be somewhat mitigated by the high radioactivity and amenability to isotopic dilution of U-233. Changes challenging Th utilization are: (a) obtaining sufficient experience handling Th/U-233 fuels, (b) the existence of large inventories of depleted U and continuing discovery of large U resources, and (c) recognition that the extent to which U-233 might mitigate proliferation concerns is not as large as originally hoped.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Decision Analysis Methods For the Analysis of Nuclear Terrorism Threats with Imperfect Information

Protection of commercial nuclear fuel cycles against adversary groups who may attempt to steal weapons-usable material is an issue of interest to the U.S. Department of Energy, Office of Nuclear Energy. The wide array of potential nuclear fuel cycle configurations and the security measures that could be deployed to protect them pose a daunting system design problem. In general, there are billions of potential design configurations. Moreover, the design choices are not independent – the effectiveness of one design feature or security measure can affect the desirability of other design choices. To explore this issue, we develop a mixed-integer optimization problem that solves a complex decision tree to allocate security resources across a commercial nuclear fuel cycle configuration. A nominal UREX+ reprocessing facility is used to demonstrate how the model allocates resources across a particular fuel cycle facility and shows how the problem can be scaled up to include other facilities and design features in the fuel cycle. In our model, the defender first chooses a set of safeguards to implement in order to mitigate the risk of a terrorist theft from various unit operations in the reprocessing facility. The attacker observes the defender’s strategy choice and executes an attack to acquire material, including the possibility of declining to attack at all. The model considers the terrorist as an intelligent adversary with a set of attack strategies known to the defender. The strategic choices made by both the defender and attacker affect the probability of a successful attack by the terrorist and in turn affect the distribution of consequences to the defender. We use a decision tree to illustrate the interactions among costs, decisions, correlated uncertainties, and outcomes, and then show how to formulate the problem as a mixed-integer program with an equivalent solution. The novel features of our formulation include the following. First, as discussed above, we use a sequential game-theoretic formulation (a Stackelberg game). This formulation assumes the attacker can observe the defender’s actions before the attack is executed. Many previous game-theoretic studies have assumed attacker and defender act simultaneously without knowledge of the other’s strategy choice. We believe the sequential game formulation is more applicable for knowledgeable insider adversaries, the focus of this study. Second, we build on our previous work that has developed a methodology to reflect the correlation among probability distributions represented as discrete chance nodes in a decision tree. We illustrate how this methodology can be applied to two chance nodes in a decision tree that relate the quality of the materials that might be stolen or diverted by an adversary with the yield of a weapon that might be constructed from that material. In a decision tree, the dependence of the yield on the quality of the material could be modeled with conditional probabilities of the weapon yield given the quality of the materials obtained. For three outcome discretizations, this would require the assessment of nine conditional probabilities from domain experts as part of the model building effort. In contrast, our methodology will allow this same representation of the tree based on the assessment of the correlation between these two chance nodes by the experts, which we believe is a much easier and more reliable assessment task. The advantage of this approach increases as the number of dependent chance nodes grows in a decision tree representation of this problem. Third, we extend this approach to the approximation of the dependence between discrete binary events such as success or failure represented by binary chance nodes in a decision tree or in the special case of a probabilistic risk analysis (PRA) based on an event tree. This extension is a novel application in this domain, and uses Monte Carlo simulation to determine the appropriate conditional probabilities on the binary chance nodes when the number of correlated uncertainties is more than two. Fourth, we show how these decision trees or probability trees can be formulated and solved as equivalent integer programming (IP) problems. The advantages of the IP representation include the opportunity to use the computing power associated with powerful optimization solver algorithms available on high speed computers to solve very large versions of these trees, and the ability to include several such models of different facilities into a much larger IP with connecting constraints that represent budget restrictions, number of available inspection opportunities in a given time window, or other dependencies among these facilities. Policy makers can use the resulting model to inform their decisions about how to assess terrorism risk and to safeguard the commercial nuclear fuel cycle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Separation of Plutonium from Irradiated Fuels and Targets

Spent nuclear fuel from power production reactors contains moderate amounts of transuranium (TRU) actinides and fission products in addition to the still slightly enriched uranium. Originally, nuclear technology was developed to chemically separate and recover fissionable plutonium from irradiated nuclear fuel for military purposes. Military plutonium separations had essentially ceased by the mid-1990s. Reprocessing, however, can serve multiple purposes, and the relative importance has changed over time. In the 1960’s the vision of the introduction of plutonium-fueled fast-neutron breeder reactors drove the civilian separation of plutonium. More recently, reprocessing has been regarded as a means to facilitate the disposal of high-level nuclear waste, and thus requires development of radically different technical approaches. In the last decade or so, the principal reason for reprocessing has shifted to spent power reactor fuel being reprocessed (1) so that unused uranium and plutonium being recycled reduce the volume, gaining some 25% to 30% more energy from the original uranium in the process and thus contributing to energy security and (2) to reduce the volume and radioactivity of the waste by recovering all long-lived actinides and fission products followed by recycling them in fast reactors where they are transmuted to short-lived fission products; this reduces the volume to about 20%, reduces the long-term radioactivity level in the high-level waste, and complicates the possibility of the plutonium being diverted from civil use – thereby increasing the proliferation resistance of the fuel cycle. In general, reprocessing schemes can be divided into two large categories: aqueous/hydrometallurgical systems, and pyrochemical/pyrometallurgical systems. Worldwide processing schemes are dominated by the aqueous (hydrometallurgical) systems. This document provides a historical review of both categories of reprocessing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

New AREVA's Investment in Continuous Improvement to Increase Efficiency of La Hague Reprocessing Lines - 18588

New AREVA currently operates 2 commercial reprocessing lines at their La Hague facility in France. Each line is capable of reprocessing up to 800 tHM/year of commercial spent light water reactor fuel. The site and process is self-contained with all waste arising from the reprocessing operations segregated into various waste streams and stabilized at the site. New AREVA operates a culture of continuous improvement and continues to invest in new technology, processes, waste management and procedures that have increased plant throughput while reducing waste volumes, operating cost while at the same time maintaining our strong safety culture. New investments have also been driven by changing customer needs and new markets. A good example of this has emerged over the last several years as New AREVA has been transporting a large amount of Research Reactor (RR) Uranium-Aluminum (UAl) type spent fuel (SF) on behalf of international RR operators to the La Hague site. Last year more than 100 RR fuel elements shipments were carried out by New AREVA (for fresh and spent fuel). In our current commercial lines the reprocessing of RRSF has been limited to approximately 1.4 ton a year. To meet increasing demand for reprocessing of RRSF, the range of RRSF that can now be reprocessed at La Hague has been extended to include Uranium-Aluminium-Silicium fuel (USi also called Silicide fuel) and also non-intact RRSF. In addition, New AREVA supported large scale research, development and engineering programs to identify processes/technology to allow the La Hague Facilities to reprocess a targeted average of additional 25 tons/year of non-commercial fuel types (RRSF and also fast reactor fuel, specific MOX fuel... ). A new facility head end called Traitement de Combustibles Particuliers (TCP) facility is currently planned to be implemented in the La Hague plant allowing reprocessed special fuels to be introduced into the existing commercial processing lines. The purpose of the paper is to demonstrate New AREVA's culture of continuous improvement that has been implemented at our commercial reprocessing facility highlighting how we have invested in processes and procedures for improving services to the nuclear operators. A case study will be presented related to our latest development of the TCP head end that will allow New AREVA to reprocess a targeted average of additional 25 tons/year of specialized fuels. (authors)

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

Adsorption of Iodine and Tritium from UNF Reprocessing Off-Gas Streams

The need to control emissions of the gaseous radionuclides to the air during nuclear fuel recycling is widely recognized. Gaseous radionuclides can partition to off-gas streams emanating from the various individual processes used in the recycle plant, such as from the mechanical head end (shear), tritium pretreatment, fuel dissolver, vessel vents, cell ventilation, and conversion operations. In traditional recycle plants, the bulk of the four primary volatile fission products (i.e., {sup 3}H, {sup 14}C, {sup 85}Kr, and {sup 129}I) is released in the head-end processes, for example, the dissolution process. The usual off-gas treatment sequence of operations targets removal of tritium (as {sup 3}HHO), iodine, carbon (as {sup 14}CO{sub 2}), and the noble gases, in that order. Water and iodine are captured first because the other species are relatively unimpeded by the sorbents used for water and iodine, so they flow downstream to additional treatment processes. Water and iodine do react with one another, and many techniques used to collect one of these species also collect some fraction of the other. It is quite desirable to sequester the iodine and water separately because the tritiated water may be disposed of as a low level waste. If not cleanly separated, contamination of the low-level waste by iodine may elevate its classification to high-level waste and result in greater costs for disposal. The proposed sorption sequence was tested in the laboratory. Use of a second column of AgZ permits the detection of breakthrough from the primary AgZ column while protecting the downstream 3AMS column from collecting iodine. (This arrangement is equivalent to enabling intra-column sampling of the primary column.) Neutron activation analysis of the 3AMS revealed that iodine was below, or at, the level of detection. The results successfully demonstrated that iodine can be sequestered in the AgZ while the water is collected free of iodine in the 3AMS column. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗