Molten salt synthesis of Ce doped zirconolite for the immobilisation of pyroprocessing wastes and separated plutonium
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99 GENERAL AND MISCELLANEOUS↗
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A Machine Learning Method for the Forensics Attribution of Separated Plutonium
Plutonium (Pu) source attribution would be a powerful tool to support nuclear nonproliferation efforts. This capability to find the source of a Pu sample would act as a deterrent to smuggling efforts, and also help regulatory agencies verify declared nuclear activities. Work at Texas A&M University yielded a nuclear forensics methodology, which is capable of determining separated Pu’s reactor of origin, fuel burnup, and the time since irradiation (TSI)—three parameters of interest. The methodology used a set of ten intra-element isotopic ratios found in separated Pu, which was compared to a library of isotopic ratio values produced using neutronics simulations for reactors of interest. By calculating the probability that unknown Pu sample’s isotopic ratio set matched a set in the library, the methodology could predict the three parameters of interest of the sample. One shortcoming of this methodology was an inability to correctly attribute spoofed Pu, where Pu sourced from two different reactors or two different fuel burnup levels are mixed. A new methodology to rectify this vulnerability using machine learning (ML) technique is developed, instead of the maximum likelihood calculation previously used and the results are satisfactory. The ML approach leverages the existing simulated data for training the algorithm, but use them efficiently by only using intra-element isotope ratios that contribute to the attribution one of the three parameters at a time. Previously, all isotope ratios were used to attribute all three parameters together. The new methodology attributes the Pu parameters in three steps, one for each parameter, rather than resolving all of the three parameters simultaneously like the previous maximum likelihood approach. First, a support vector machine classifier with a set of seven isotopic ratios finds the reactor of origin and a set of regression models trained using gaussian process predicts the burnup with a different set of seven isotopes. Finally, TSI is calculated analytically using decay equations. Thus far, the new methodology is capable of attributing pure Pu samples and has been validated using experimental data. The next step will to be augment the classifier training data set with spoofed Pu data.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Validation of a Machine Learning–Based Nuclear Forensics Methodology for the Discrimination of a Chemically Separated Plutonium Sample from Low-Enriched Uranium
Not provided.
Nuclear Science & Technology↗
Utilization of traceable standards to validate plutonium isotopic purification and separation of plutonium progeny using AG MP-1M resin for nuclear forensic investigations
Radio-chronometric studies on plutonium (Pu) materials require independent measurement of the Pu (parent) content and isotopic distribution as well as concentration and isotopic distribution of the plutonium isotopic decay products. We performed a series of experiments to demonstrate the consistency of separations using the Lewatit MP 800 macroporous anion exchange resin and the AG MP-1M resin with traceable Pu isotopic certified reference material (CRM) standards 136, 137, 138, and 126-A. Two different mesh-sizes of the AG MP-1M resin were tested and the 50–100 mesh size resin was found to work more efficiently for the separation task. Both Lewatit and AG MP-1M resins were found to perform satisfactorily for quantitatively extracting the americium (Am) and uranium (U) progeny as well as gallium (Ga) present as a tracer in the Pu material. Both resins were effective in removing isobaric interferences from the Pu fraction used in isotopic measurements by thermal ionization mass spectrometry (TIMS). To address the co-elution of uranium and gallium, Alizarin red S (ARS) was used as a colorimetric dye to determine the behavior of UO 2 2+ and Ga 3+ on AG MP-1M resin with various acidic solutions as eluents using UV–vis spectra. Poor resolution of these peaks complicated quantitative analysis by UV–vis spectroscopy, but these results were informative in planning automated separation experiments by HPLC. LA-UR-24-28919.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Photochemical separation of plutonium from uranium
Photochemical reduction and separation of plutonium from uranium in acidic solutions is described as a potential alternative to conventional separations that employ harsh chemical redox agents.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Separation of dilute plutonium(IV) from nitric acid solutions via Reillex HPQ anion exchange resin
Anion exchange is a common method for the separation of plutonium (Pu) in aqueous systems, providing high recovery and decontamination from impurities. Reillex HPQ anion exchange resin has been used around the Department of Energy complex in this application for decades but has typically been used for Pu feed concentrations greater than 1 g/L. It is desirable to recover Pu at more dilute concentrations from a dissolved used fuel or target element, for example. In this work, the performance of Reillex HPQ anion exchange resin was characterized with feed solutions of ca. 0.24 g/L Pu. A significant reduction in the performance of the resin was realized with this dilute concentration relative to a feed solution of 2.35 g/L Pu with similar feed flow rates and the same resin column, likely due to the increased driving force for mass transfer with a higher feed concentration. The maximum capacity of Reillex HPQ resin for Pu can be improved by reducing the flow rate and increasing the feed Pu concentration. Furthermore, the breakthrough point of Pu can be improved in the same manner and by increasing the resin bed height. Diffusion of Pu into the resin was illustrated to be the rate limiting step.
Plutonium separations↗
Effect of ionizing radiation on PuO2 and surrogate oxide materials under storage conditions
The stockpile of plutonium separated from used nuclear fuel amounts to many 100s of tonnes. The majority of this plutonium is stored as PuO2 powder packaged in multi-can containers. The conditions during the packaging of reprocessed PuO2 for long term storage are closely controlled to limit water uptake as radiolytic decomposition of any adsorbed water will lead to the formation of both a potentially flammable atmosphere containing molecular hydrogen as well as reactive oxygen species which may be incorporated into the oxide phase. Consequently, the safety case for the long-term storage of PuO2 requires a complete understanding of the fundamental physical, chemical and materials degradation processes occurring at the water-oxide interface inside the storage canisters. This study investigates the effect of ionizing radiation on PuO2 and various surrogate oxides under atmospheres potentially encountered in storage canisters.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗
Chemical processes for recovery and purification of high-purity uranium-234 from aged plutonium-238
This work describes a streamlined approach to the recovery and purification of 234 U from aged 238 Pu. Key modifications to conventional techniques are described that result in an efficient uranium and plutonium separation process that successfully produced tens of grams of ultrahigh purity 234 U at Oak Ridge National Laboratory. A new processing scheme was developed using an anion exchange column run in a cycling mode with HNO 3 media as the primary uranium purification process. This was followed with a uranyl peroxide precipitation step to purify and solidify the uranium before conversion to U 3 O 8 . Both the traditional uranyl peroxide precipitation method and a new cascade uranyl peroxide precipitation process were used for final purification of the 234 U.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
The Jupiter Experiments: High-240 Plutonium Metal Plates Separated by Lead and Reflected by Copper
Each layer consisted of a 6 by 6 matrix of either these fuel-filled containers or solid blocks of copper of the same outer dimensions. An example of this arrangement, along with one of the copper inner reflectors, is shown in Figure 1.4. This figure shows the same aluminum containers from Figure 1.3 but with the aluminum containers completely closed. The lifting rings shown on the copper inner reflector were only for assembly and were not present for the measurement (in which the holes were filled with copper plugs). Figure 1.5 shows the loading arrangement of the aluminum containers, along with aluminum shims around the container arrays to ensure a tight fit between these layers and the surrounding reflectors. All plates were in the same orientation; none were rotated in any fashion. Additional loading information for the PAHN plates and plate loading respective to the room entrance are provided in Figures 1.6 and 1.7, respectively.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Nuclear Weapons Production – Materials Cycle [Slides]
Presentation Overview: This presentation will familiarize participants with the production of plutonium, its separation, conversion to metal, casting, machining, and assembly of special nuclear materials common to nuclear weapons, and the disposal of nuclear waste. Focus in this talk is plutonium (in future briefings we will broaden to include other materials).
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
An approach to separating Pu, U, and Ti from high-purity graphite for isotopic analysis by MC-ICP-MS
Information about elemental and isotopic systematics of ultra-trace level actinides (e.g. U and Pu) and main group elements (e.g. Ti) present within nuclear grade graphite is vital to the nuclear community for improved reactor operation and security. In support of this, extensive effort has been placed on improving analysis methods (i.e., inductively coupled plasma-mass spectrometry). However, significantly less effort has been devoted to the optimization of chemical separation methods. Within the separation community, commercially available Eichrom™ resins are often employed, as their elution characteristics for various elements have been well studied, but the direct optimization of actinides and trace metal separations from a single sample have not been widely investigated. Here, methods using various Eichrom pre-packed cartridges were explored to achieve separation of ultra-trace levels of U, Pu, and Ti from a variety of graphite samples. Once the validity of the combined separation scheme was established using certified reference materials, the method was applied to historic, unirradiated and irradiated, graphite samples. For all samples investigated, precise isotope ratio measurements for the titanium isotope systems were made.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Transplutonium elements processed from rock debris of underground detonations
Six-step chemical processing method extracts minute quantities of transplutonium elements found in rock debris following a nuclear detonation. The process consists of dissolution of rock, feed preparation, liquid-liquid extraction, final purification of transplutonium elements and plutonium, and separation of the transplutonium elements.
Bloomquist, C. A. A.↗
Review of historical monitoring data on Techa River contamination
The Mayak Production Association was the first Russian site for the production and separation of plutonium. The extensive increase in plutonium production during 1948-1955, as well as the absence of reliable waste-management technology, resulted in significant releases of liquid radioactive effluent into the rather small Techa River. This resulted in chronic external and internal exposure of about 30,000 residents of riverside communities; these residents form the cohort of an epidemiologic investigation. Analysis of the available historical monitoring data indicates that the following reliable data sets can be used for reconstruction of doses received during the early periods of operation of the Mayak Production Association: Temporal pattern of specific beta activity of river water for several sites in the upper Techa region since July 1951; average annual values of specific beta activity of river water and bottom sediments as a function of downstream distance for the whole river since 1951; external gamma-exposure rates near the shoreline as a function of downstream distance for the whole Techa River since 1952; and external gamma-exposure rate as a function of distance from the shoreline for several sites in the upper and middle Techa since 1951.
NASA Discipline Radiation Health↗
Lattice Confinement Fusion and Fusion-Fast-Fission Energy Source Development
NASA and the US Navy have demonstrated Lattice Confinement Fusion (LCF) and the Fusion-Fast-Fission of natural uranium and thorium. Both methods build upon decades of research, and the latter benefits from conventional nuclear fission and fusion results. However, this nascent technology doesn’t require fissile isotopes and avoids nuclear weapons proliferation concerns from uranium 235U isotopic enrichment or plutonium 239Pu separation. Of particular note, LCF doesn’t need power hungry magnets, lasers, or particle beams. Potentially, it could provide watts to hundreds of kilowatts of electrical power and process heat suitable for space power and high Isp nuclear electric propulsion (NEP) as well as distributed terrestrial power. Finally, there are indications that LCF fast-fission products are more benign than those of conventional fission reactors. Our goal is to scale the reactions and increase the power output through higher temperature operation and increased material masses.
lattice confinement fusion↗
Multiscale Evaluation of Acetohydroxamic Acid (AHA) Radiolysis Under Used Nuclear Fuel Reprocessing Solvent System Conditions
Acetohydroxamic acid (AHA) has been proposed as an alternative agent for the selective separation of plutonium and neptunium from co-extracted uranium during the reprocessing of used nuclear fuel. However, the fundamental radiolytic behavior of this molecule under envisioned process conditions – i.e., acidic biphasic solvent systems – is not sufficiently understood to support process applications. Here we present a systematic irradiation study (steady-state gamma and time-resolved pulsed electron) into the radiolytic integrity of AHA and formation of degradation products in aqueous nitric acid (HNO3) solutions (0.2 M) in presence and absence of an organic phase, comprising current (tri-butyl phosphate - TBP) and future (N,N-di-(2-ethylhexyl)butyramide - DEHBA and di-2-ethylhexylisobutyramide - DEHiBA) reprocessing ligands dissolved in n-dodecane diluent. Experimental data are complimented by predictive multiscale model calculations for the elucidation of underpinning mechanisms.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Comparison of Industrialized Late 20th Century Flowsheets for Reprocessing Used Nuclear Fuel
This study identified and compared three flowsheets for reprocessing used nuclear fuel (UNF) industrialized in plants in the United Kingdom (the Thermal Oxide Reprocessing Plant), France (UP2-800/UP3) and Japan (the Rokkasho Reprocessing Plant). The study also identified the major implications for a plant in the United States if it were initiated. All flowsheets employed the established Plutonium Uranium Reduction Extraction (PUREX) solvent extraction technology to separate uranium and plutonium from UNF dissolved in nitric acid. However, differences in the approaches to managing iodine-129, tritium and technetium were identified in the flowsheets. A US plant would also need to separate krypton-85 as well as iodine-129 and tritium for immobilization and disposal.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Unravelling the radiation-induced redox chemistry of plutonium ions in aqueous solution
Plutonium plays a critical role in nuclear fuel cycle technologies, but our understanding of its fundamental radiation-induced redox chemistry is limited. Changes in oxidation states affect the speciation and transport of plutonium ions in solution. For example, solvent extraction techniques used to separate and recover plutonium from used nuclear fuel rely on the selective formation, maintenance, and complexation of specific plutonium oxidation states. However, radiolytically generated radicals, ions, and molecules can drive the oxidation state distribution of plutonium ions far from equilibrium, ultimately changing the physical and chemical properties of the bulk system. These radiation-induced processes are inevitable due to the ionizing radiation fields generated by the radioactive decay of plutonium and its daughter nuclides. Therefore, mechanistically understanding how plutonium's various oxidation states respond to ionizing radiation is essential for predicting its behavior in solution. Here, we present significant advances in our understanding of radiation-induced plutonium redox chemistry by using time-resolved (electron pulse) and dose accumulation (alpha and gamma) irradiation techniques, along with quantitative multiscale modeling methods.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗