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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Compositions and methods for monitoring actinides

Compositions and methods for monitoring the quantity of actinides present in a test sample are disclosed. Compositions and methods for monitoring the motion of special nuclear materials through space are also described. Compositions and methods for monitoring the quantity of a fissile special nuclear material present in a test sample are disclosed. Compositions and methods for monitoring actinides during reprocessing of spent nuclear fuel after 30-year cool down are disclosed. Compositions and methods for monitoring actinides during reprocessing of spent nuclear fuel after 180 day cool down are also disclosed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Researching the Effects of Long-Term Dry Storage on Commercial Spent Nuclear Fuel

When the fuel in a nuclear reactor no longer contributes efficiently to power production, it is discharged and stored onsite in a water-filled spent fuel pool. When nuclear power plants in the United States were originally designed, it was expected that onsite storage would be temporary. The used nuclear fuel, also known as spent nuclear fuel (SNF), was to be removed from the storage pools after only a few years of cooling and sent for reprocessing. However, commercial reprocessing was not implemented in the United States, and utility spent fuel pools began running out of room. In order to continue power production, nuclear utilities began expanding onsite fuel storage capacities, first by increasing the density of fuel in the fuel pool racks, and then by developing and implementing dry storage casks. The dry storage casks are stored onsite by each utility at a licensed and secured independent spent fuel storage installation (ISFSI). Dry storage casks are robust systems typically consisting of a sealed metal cylinder surrounded by a combination of steel and concrete to provide radiation shielding. The first dry storage casks were loaded with SNF in 1986. Currently there are over 3,000 loaded dry storage casks, and approximately 200 new casks are being placed into service annually at operating nuclear power plants nationwide. Multiple nuclear power plant sites in the United States have shut down and completed decommissioning activities except for the removal of fuel in dry storage. Eventually, the 75,000+ tons of discharged SNF at more than 70 secured locations will need to be safely and securely transported to a new location(s) for eventual disposition. The continued safety and security of this material must be assured.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

The MPACT 2020 Milestone: Safeguards and Security by Design of Future Nuclear Fuel Cycle Facilities.

The Materials Protection, Accounting, and Control Technologies (MPACT) campaign, within the U.S. Department of Energy Office of Nuclear Energy, has developed a Virtual Facility Distributed Test Bed for safeguards and security design for future nuclear fuel cycle facilities. The purpose of the Virtual Test Bed is to bring together experimental and modeling capabilities across the U.S. national laboratory and university complex to provide a one-stop-shop for advanced Safeguards and Security by Design (SSBD). Experimental testing alone of safeguards and security technologies would be cost prohibitive, but testbeds and laboratory processing facilities with safeguards measurement opportunities, coupled with modeling and simulation, provide the ability to generate modern, efficient safeguards and security systems for new facilities. This Virtual Test Bed concept has been demonstrated using a generic electrochemical reprocessing facility as an example, but the concept can be extended to other facilities. While much of the recent work in the MPACT program has focused on electrochemical safeguards and security technologies, the laboratory capabilities have been applied to other facilities in the past (including aqueous reprocessing, fuel fabrication, and molten salt reactors as examples). This paper provides an overview of the Virtual Test Bed concept, a description of the design process, and a baseline safeguards and security design for the example facility. Parallel papers in this issue go into more detail on the various technologies, experimental testing, modeling capabilities, and performance testing.

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

Safeguards Modeling for Advanced Nuclear Facility Design.

Future nuclear fuel cycle facilities will see a significant benefit from considering materials accountancy requirements early in the design process. The Material Protection, Accounting, and Control Technologies (MPACT) working group is demonstrating Safeguards and Security by Design (SSBD) for a notional electrochemical reprocessing facility as part of a 2020 Milestone. The idea behind SSBD is to consider regulatory requirements early in the design process to provide more optimized systems and avoid costly retrofits later in the design process. Safeguards modeling, using single analyst tools, allows the designer to efficiently consider materials accountancy approaches that meet regulatory requirements. However, safeguards modeling also allows the facility designer to go beyond current regulations and work toward accountancy designs with rapid response and lower thresholds for detection of anomalies. This type of modeling enables new safeguards approaches and may inform future regulatory changes. The Separation and Safeguards Performance Model (SSPM) has been used for materials accountancy system design and analysis. This paper steps through the process of designing a Material Control and Accountancy (MC&A) system, presents the baseline system design for an electrochemical reprocessing facility, and provides performance metrics from the modeling analysis. The most critical measurements in the electrochemical facility are the spent fuel input, electrorefiner salt, and U/TRU product output measurements. Finally, material loss scenario analysis found that measurement uncertainties (relative standard deviations) for Pu would need to be at 1% (random and systematic error components) or better in order to meet domestic detection goals or as high as 3% in order to meet international detection goals, based on a 100 metric ton per year plant size.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2014

A dark matter search using the final CDMS-II data and 100 mm SuperCDMS germanium detector ionization test

Astrophysical observations indicate that approximately 85% of the matter in the universe is nonluminous, nonbaryonic, and nonrelativistic (cold) dark matter. Weakly Interacting Massive Particles (WIMPs) are a particularly well motivated dark matter particle candidate. They would be thermally produced in the early universe and their relics account for the current dark matter abundance. WIMP candidate particles are naturally provided by extensions to the Standard Model of particle physics, such as supersymmetry. The Cryogenic Dark Matter Search (CDMS) experiment operates cryogenic germanium and silicon particle detectors in the low-background environment of the Soudan Underground Laboratory in northern Minnesota to search for WIMP-nucleus scatters while rejecting electron-recoil background. The detectors simultaneously measure the ionization and phonon energies of each scattering event. The difference in ionization yield (ratio of ionization energy to recoil energy) discriminates nuclear recoils from the electron-recoil background efficiently.More sensitive detectors are required to probe the WIMP parameter space with lower WIMP-nucleon scattering cross sections. To support the R&D effort especially the detector R&D and characterization of the SuperCDMS experiment, a new CDMS test facility has been developed on the University of Minnesota campus. This thesis documents the test facility and the work involved in its development. In the test facility, we performed the first ionization collection efficiency measurements of the ionization test devices. The test devices are fabricated with detector-grade germanium crystals that are 100 mm in diameter, which is the largest available, and 33 mm in thickness. The measured efficiencies are consistent with the earlier measurements conducted with smaller Ge crystals, demonstrating that these 100 mm crystals can be used for development of the next generation dark matter detectors.Improvements of data analysis methods can also potentially improve the sensitivity of an experiment. The data taken during the last four runs of CDMS II with total raw exposure 612 kg-day were reprocessed with improved ionization pulse reconstruction algorithm. We present the classic timing analysis with the reprocessed data in this thesis. For the four runs combined, this analysis resulted in a new WIMP-nucleon cross section 4.4×10 -44 cm 2 for a WIMP mass of 70 GeV, which is a factor of 1.6 improvement compared to the original c58 classic timing analysis.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2023

Integration of Nuclear Material Accounting Data and Process Monitoring Data for Improvement on Detection Probability in Safeguarding Electrochemical Processing Facilities (Final Technical Report)

The KAERI advanced spent fuel conditioning process (ACP) process is a critical component of the US- South Korean nuclear cooperation and the following “123 Agreement.” Its development has received considerable attention in both countries. The ACP is an electrochemical processing (pyroprocessing) that recycles over 96% of the used nuclear fuel (UNF). It is also intrinsically proliferation-resistant in theory. In normal operation, the U/TRU product is very hot radiologically. In addition, the Cm provides a high level of spontaneous neutrons, making the product unsuitable for weapon use. However, as pointed in some study, “the need for safeguards to protect against the diversion and misuse of separated plutonium applies essentially equally to all grades of plutonium.” As pointed by many studies, the well-established traditional Nuclear Material Accounting (NMA) approach cannot be directly applied to electrochemical processing because of the lack of an input accountability tank, the non-continuous material flow, and the unsatisfactory level of confidence in sampling methods. Therefore, nuclear safeguards remain a grand challenge in the developing of commercial electrochemical separations facilities, especially around the heart of such facilities, the electrorefiner (ER) systems. In contrast to NMA data, process monitoring (PM) data is normally an indirect measurement of the SNM and is acquired much more frequently. In a broad sense, PM includes monitoring by various types of equipment, e.g. radiation detectors, cameras, voltage, current sensors. Because it is already being collected by the operator, the additional cost to safeguards is low. It has long been believed that PM data can supplement NMA data and help improve safeguards, although the benefits are hard to quantify. The U.S. DOE’s Material Protection, Accounting, and Control Technology (MPACT) campaign has made substantial investments into innovative PM sensor technology and predictive model development for real- or near real-time measurement and prediction of molten salt density and level, salt composition and actinide concentration especially Pu, the cell voltage, and the cell current to supplement traditional NMA. For aqueous-based reprocessing facilities, it is reported that PM, integrated with traditional NMA, have a high detection probability for specific diversions. For electrochemical reprocessing, preliminary studies have shown that PM data can support traditional NMA in various ways by providing a basis to estimate some of the in-processing nuclear material inventories. Despite early success, further studies on fusion of PM data and NMA data are still needed, which is the goal of this proposed work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Radiolytic transformation of AHA under single-cycle conditions

Historically, the radiation robustness of extractant ligands and the impact of their radiolysis products have been used to evaluate the radiolytic feasibility of solvent system formulations for used nuclear fuel (UNF) reprocessing. However, other key additives (e.g., reducing and hold-back reagents) are employed by these solvent systems. These chemical species have typically been overlooked from a radiolytic perspective, and consequently little is known on how they impact process performance. Therefore, to holistically support the development of advanced, simplified, single-cycle flowsheets for UNF reprocessing, the radiation robustness of two key proposed additives – (i) the complexing reductant acetohydroxamic acid (AHA) and (ii) the hold-back reagent 1,2-cyclohexylenedinitrilotetraacetic acid (CDTA) – have been evaluated under envisioned process conditions using a combination of steady-state gamma and time-resolved pulsed electron irradiation techniques.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Enabling Recycling of Composites: Understanding the Impacts of Multiple Thermal Processing Cycles

When considering the utilization of recycled short carbon fiber feedstock materials for advanced manufacturing, understanding the material degradation behavior is essential in determining how many times a composite material can be effectively reprocessed and remanufactured. This study characterizes the degradation behavior of short carbon fiber acrylonitrile butadiene (CF-ABS) that has been reprocessed five times with twin screw extrusion. Parallel plate rheology was completed to observe the degradation in complex viscosity of the recycled feedstock materials. Gel permeation chromatography (GPC) was utilized to characterize the changes in molecular weight distribution of the recycled materials as a result of thermal and mechanical degradation during the re-processing steps. Rheological characterization, GPC, and twin-screw processing data help inform the process optimizations required to process the recycled feedstock material. Successful characterization of the degradation behavior of short fiber composite feedstock materials aids in increased understanding of the lifespan of high value carbon fiber composite materials and aids in process optimization of recycled composite materials.

Walker, Roo↗
DOE OSTI2015

Improving actinides recycling, a key step towards a more sustainable nuclear energy - 15617

Nuclear energy is thought to be one of the energy sources that could help mitigating the global climate change together with the renewables, due to its low green-house-gases emissions, its reliability and its high load power. However, nuclear energy will only develop if future nuclear energy systems meet the criteria of sustainability in terms of durability, bearability and liveability. Nowadays, many countries chose the so-called once-through cycle which basically considers spent nuclear fuel as a waste, whereas others reprocess their spent fuel to recover the energetically-valuable material Pu (and partially U) to produce Mixed Oxide Fuel (MOX) to be irradiated in a second cycle (twice-through cycle). Both options can be considered relevant, fitting diverse criteria; but if we consider a long lasting prospective for nuclear options, the once-through cycle sustainability is clearly lower (low consumption of natural resource, no recycling of potentially valuable energetic material, larger volume of waste...). Reversely, twice-through cycle can be seen as a first step towards an improved sustainability with a partial recycling of U/Pu and the reduction of the waste volume and toxicity. However, it is likely not sufficient for the long-term and implementing more efficient recycling is a key-driver to improve the environmental friendship, the economics and the social acceptability of nuclear energy. Future nuclear fuel cycles will have to better preserve the natural resource by improving the actinides recycling, to minimize the volume and toxicity of nuclear waste, to ensure their long-term confinement and to prevent any proliferation-risk. Such a gradual and progressive process calls for implementing appropriate technologies, either for reactors but also for fuel treatment/recycling facilities. This paper will describe the successive step that would allow improving the overall sustainability. Indeed, recycled Pu can be once recycled in MOX fuel feeding similar PWRs (Pressurized Water Reactor). Together with the recycling of reprocessed uranium, this first step already allows preserving ∼17% of the natural U resource and to confine ultimate wastes without Pu in a tailored durable wasteform. Subsequent Pu multi-recycling requires introducing Fast Reactors (FR) to efficiently use U-238 available in natural and depleted uranium stockpiles. In parallel, recycling processes would have to be adapted for treating spent LWR and SFR MOX spent fuels, which is currently under study thanks to an ambitious R and D program dedicated on MOX treatment and recycling. For the longer term, minor actinides recycling would allow decreasing the waste burden towards future generations: (i) Am-recycling leads to decrease the waste residual heat-power and increase the repository density and lifespan, (ii) Cm-recycling leads to decrease waste long-term toxicity. The stepwise implementation of these different processes defines a path towards sustainability for nuclear energy that are described in this paper from the fuel cycle viewpoint. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Waste Management Approach for a Potential Future U.S. Closed Fuel Cycle - 16233

The U.S. Department of Energy, Office of Nuclear Energy has chartered an effort to develop technologies to enable safe and cost effective recycle of commercial used nuclear fuel (UNF) in the U.S. Part of this effort includes the evaluation of exiting waste management technologies for effective treatment of wastes in the context of current U.S. regulations and development of waste forms and processes with significant cost and/or performance benefits over those existing. This study summarizes the results of these ongoing efforts with a focus on the highly radioactive primary waste streams. The primary streams considered and the recommended waste forms include: - Tritium separated from either a low volume gas stream or a high volume water stream. The recommended waste form is low-water cement in high integrity containers. - Iodine-129 separated from off-gas streams in aqueous processing. There are a range of potentially suitable waste forms. As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals. - Carbon-14 separated from LWR fuel treatment off-gases and immobilized as a CaCO{sub 3} in a cement waste form. - Krypton-85 separated from LWR and SFR fuel treatment off-gases and stored as a compressed gas. - An aqueous reprocessing high-level waste (HLW) raffinate waste which is immobilized by the vitrification process in one of three forms: a single phase borosilicate glass, a borosilicate based glass ceramic, or a multi-phased titanate ceramic [e.g., synthetic rock (Synroc)]. - An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel that is either included in the borosilicate HLW glass or is immobilized in the form of a metal alloy in the case of glass ceramics or titanate ceramics. - Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware that are washed and super-compacted for disposal or as an alternative Zr purification and reuse (or disposal as low-level waste, LLW) by reactive gas separations. - Electrochemical process salt HLW which is immobilized in a glass bonded Sodalite waste form known as the ceramic waste form (CWF). - Electrochemical process UDS and SS cladding hulls which are melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. (authors)

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

UMo Solutions Processing in La Hague Cold Crucible Induction Melter: the Feedback From the First Years of Operation - 16376

250 cubic meters of high-level liquid waste from reprocessed U-Mo-Sn-Al spent fuel, used in Gas Cooled Reactors (GCR), were produced during the mid-1960's at La Hague facility. These 'UMo' solutions are less radioactive than the current fission product concentrates coming from ongoing reprocessing activities, but are very rich in molybdenum and phosphorus whose contents make the molten glass quite corrosive and require a high-temperature glass formulation to obtain sufficiently high waste loading factors (12% in molybdenum oxide). Hence the use of the Cold Crucible Induction Melter (CCIM) technology to process such solutions has been deemed a good opportunity for AREVA to meet its performance expectations. In addition to being very corrosive, the UMo waste is quite challenging to process as the molybdenum has a strong tendency to stick in the calciner and causes clogging issues in off-gas treatment equipment. Therefore, the process and technological qualifications were deployed in order to address these specific issues. UMo solutions processing in the La Hague CCIM has started in January 2013 and is currently ongoing. During this period (from 2013 to 2015), many data have been collected to confirm the process parameters that were defined during the qualification of this innovating process. Even if some difficulties occurred, operations teams experience along with engineering and R and D support allowed managing them. This paper presents the start-up methodology and the feedback from the first years of UMo solutions processing with the CCIM technology at La Hague site. Lessons learned are presented with the difficulties encountered and the solutions implemented, emphasizing the benefits of a close integration between R and D, engineering and operations teams. (authors)

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

Choosing Your Nuclear Fuel Cycle: A Life Cycle Assessment Perspective - 16425

Nuclear waste management has always been a much debated topic and a key driver for decisions within the nuclear industry. A standardised global approach has not yet been developed and at present many countries are re-considering their position. In the UK the reprocessing of Spent Nuclear Fuel is projected to stop, but no clear indication about the future approach has been given. This contribution sets out the early stages of an approach to compare the possible alternatives based on Life Cycle Assessment (LCA), which is widely used in other sectors to assess the overall environmental burdens of a product or a service across the whole life cycle - i.e. from cradle to grave. The LCA may serve to support decision-making processes within the nuclear industry and, provided that it is used in an open and transparent way, to improve public attitudes towards nuclear energy. To date few LCA studies within the nuclear industry have been carried out, the main issue being the lack of a standardised methodology to evaluate the impacts of radionuclides. The present study proposes a new framework for this purpose, and demonstrates its application in an LCA study of the current UK approach to the management of spent nuclear fuel based on specific data from plants at Sellafield. The final results of the study will show the overall environmental footprint of the reprocessing of spent nuclear fuel, in terms of both radiological and non-radiological impacts. A 'hot spot' analysis will also be performed to highlight the critical processing steps. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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 OSTI2016

Separation of Zirconium from Zircaloy-2 for Spent Nuclear Fuel Cladding Applications using a Chloride Volatility Process

The United States produces a large amount of cladding material from used nuclear fuel, expected to approach 1,000 MT/year for the next 50 years. The recovery of zirconium from the cladding material would represent a ∼50% savings on the geologic waste disposal requirement, and savings of >$40 M/year due to the cost of hafnium-free zirconium. Previous simulation studies have shown that in the irradiated cladding, the major activation products found are {sup 125}Sb, {sup 60}Co, and {sup 93m}Nb, and the major fission product found is {sup 137}Cs. These radioisotopes make it impossible to treat the cladding as low level waste. In a chlorination method, zirconium cladding reacts with Cl{sub 2} at 350-380 deg. C to produce mainly ZrCl{sub 4}. The zirconium tetrachloride has a relatively low boiling point (sublimes at 330 deg. C) which allows its separation from other products, and so it can be treated as low level waste for disposal or recycling. This chlorination approach has been proposed for the reprocessing of zirconium. However, previous work has observed element impurities (Fe, Sn, Cr, Cs, Sb, and Nb) in the recovered ZrCl{sub 4} meaning that additional steps are necessary in a chlorination process previous to the reduction of ZrCl{sub 4} to the metallic form. The chlorination process for zirconium reprocessing has been highly encouraged as it has proven more efficient than previous mechanical processes, redox processes, and other chemical methods such as iodination. The overall goal of this project is to develop the road map for a chlorination process that will yield high purity zirconium. The current study focuses on the recovery of ZrCl{sub 4} from non-irradiated alloys used as nuclear fuel cladding. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Analysis of Synergistic Fuel Cycle Options with Thorium and Heavy Water Reactors

Fuel cycle options (FCOs) with thorium-based fuels have characteristics that may have benefits for meeting certain objectives, such as transuranic actinide inventory management and overall natural resource utilization. Neutron irradiation of thorium and uranium-233 results in negligible production of transuranic actinides. This advantage is compromised, to a degree, when thorium fuels are used in conjunction with uranium-238; however, strategic reprocessing and fuel use (e.g., recycle of plutonium and/or minor actinides) can be used to manage this problem. Furthermore, thorium FCOs with reprocessing can extend the availability of natural nuclear resources by breeding new fissile material, recovering and recycling unused material, and achieving high burnups. In instances where natural thorium is more readily available than uranium, FCOs which use thorium may be preferable from sustainability and energy security perspectives. However, many hybrid uranium-thorium FCOs struggle to perform well in this regard, since large amounts of external fissile material (and therefore natural uranium) are required to breed the required uranium-233. DOE's evaluation and screening effort reports that for every metric ton (MT) of U-233 bred in a light water reactor for reuse, about 1000 MT of natural uranium are required to support such breeding, much of which is used inefficiently due to enrichment. The combination of HWRs and thorium-based fuels has the potential to reduce this consumption of natural uranium resources. Because heavy water moderates neutrons without significant absorption, fissile material requirements are lower compared to those of LWRs. This has enabled conventional 'once-through' HWRs to operate with only slightly-enriched or even natural uranium. Also, HWRs can be combined with LWRs in multistage FCOs where used LWR fuel can be processed and the remaining fissile content is sufficient to support use as fuel in HWRs in spite of the presence of neutron poisons (e.g., uranium-236); this concept has been demonstrated at a research scale in Korea. Thus, additional energy is extracted from the same initial natural uranium feed and uranium enrichment effort. Another consideration with using thorium-based fuels in HWRs is the current limited operational experience of the reactors and associated fuel cycle facilities using these fuels. However, work continues internationally towards gaining further experience. For example, at equilibrium the third stage of India's planned Three-Stage Nuclear Power Program uses thorium/uranium-233 or thorium/plutonium fuels with a fissile supply from second-stage fast reactors. Further, China plans to experiment with low-enriched uranium (LEU)-Th CANFLEX fuel bundles in a once-through configuration in its HWRs and a long-term objective is to transition to thorium use in a closed HWR FCO. Vanderbilt University (VU) and Oak Ridge National Laboratory (ORNL) are characterizing multi-stage, thermal-reactor FCOs that incorporate thorium and examining their potential performance. This paper presents insights from the examination of FCOs which consider the synergy of HWRs and thorium. Two multi-stage FCOs which incorporate thorium-based fuels in HWRs have been examined for their potential, one which uses recycled PWR uranium oxide fuel in conjunction with thorium in an HWR, and another that uses recycled HWR fuel in an MSR which also includes thorium. Both of the systems studied offer resource utilization improvements as compared to a once-through LWR fueled with low-enrichment uranium and highlight the potential synergy of thorium and HWRs. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Effect of Moisture on the Adsorption of Iodine and Krypton in Multicomponent Streams on the 10 WT% C/ETS-10 Sorbent

Over 13% of the world's energy demand is supplied by nuclear power plants in 31 countries. This demand is growing; by 2030 the IAEA predicts a 100% increase in the demand for nuclear energy. The increase in nuclear energy production imposes a growing strain on the uranium fuel supply. At current consumption rates the world's supply of uranium is projected to last ∼200 years. With the growing energy consumption, uranium supplies will dwindle even more quickly. However, ∼95% of uranium is left in used nuclear fuel (UNF) and may be utilized through reprocessing. A main concern with reprocessing is the release of off-gas. The major constituents of off-gas include {sup 85}Kr, {sup 129}I, {sup 14}C, {sup 3}H, and various Xe isotopes. {sup 85}Kr and {sup 129}I present the largest threat due to their harmful environmental effects and abundance. Several methods have been studied for the removal of these contaminants, including cryogenic distillation, fluorocarbon absorption, wet scrubbing, and adsorption have been employed. Adsorption on porous solid sorbents has the advantages of being simple, effective, and economic. Thus many studies have been performed investigating several sorbents such as silver impregnated mordenite, silver Engelhard titanosilicate 10, chalcogen-based aerogels, and metal organic frameworks. Despite these advances, many of these studies have been conducted under ideal conditions: cryogenic temperature, helium as a carrier gas, and single component adsorption. The need for investigating adsorption under conditions more representative of practical use is needed. Thus our work has been devoted to the study of adsorption of {sup 85}Kr and {sup 129}I from a simulated off-gas on an Engelhard titanosilicate 10 (ETS-10) supported carbon nanopolyhedron sorbent (C/ETS-10) in a multicomponent system at ambient temperatures. Previously we have reported sorption capacities of C/ETS-10 for single component krypton, and iodine. The effects multi-contaminant iodine and krypton and their respective inlet concentrations on the sorption capacity of 10 wt% C/ETS-10 have also been studied. Investigations into the effect of the presence of moisture in the multicomponent gas mixture on the sorption on 10 wt% C/ETS-10 are described in this paper.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Solids from Fuel Dissolution, Process Solutions, and Waste Evaporation

Irradiated light water reactor (LWR) fuel contains material which is insoluble in nitric acid, and process solutions are unstable with respect to precipitation of compounds insoluble in nitric acid. Accordingly, solids will be encountered during fuel reprocessing, and operations must be conducted in a way to minimize formation of additional solids and to prevent their interference with reprocessing operations.

Campbell, D. O.↗
DOE OSTI2024

Investigating the Impacts of Direct Dissolution Conditions on the Radiolytic Longevity of Butyramide Extractants

Removing the nitric acid (HNO3) dissolution step in used nuclear fuel (UNF) reprocessing would reduce the volume of radioactive waste streams generated, thereby, improving process efficiency. A promising strategy for this is the direct dissolution of UNF that has been pretreated by voloxidation into an organic solvent composed of specialized extractants and diluent. However, removal of the aqueous HNO3 phase from the envisioned reprocessing system has the potential to drastically change the suite of radiation-induced processes occurring, and thus, alter the longevity of proposed reagents. Furthermore, the impacts of fission product and transuranic metal ion complexation on the aforementioned radiation-induced processes is poorly understood, and yet can cause significant changes in radiolytic longevity. To bridge these knowledge gaps and support the continued development of direct dissolution strategies, we present an investigation into the impacts of direct dissolution conditions on the gamma radiation-induced degradation of N,N-di-(2-ethylhexyl) butyramide (DEHBA) and N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) ligands—candidate replacements for tributyl phosphate—in pre-equilibrated n-dodecane solvent in the presence and absence of envisioned loading amounts of uranium.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗