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At least 631 records · Page 35

Equity Considerations in Siting Consolidated Interim Storage Facilities for Commercial Spent Nuclear Fuel

Questions of equity and fairness are often raised by representatives from State, Tribal and local governments and members of the public when resolving conflicts over siting of nuclear facilities. However, the two terms are not synonymous: equity can be quantified while fairness is much more subjective. The Department of Energy's (DOE) December 2021 Request for Information sought input on how to site federal facilities for the temporary, consolidated storage of spent nuclear fuel (SNF) using a consent-based approach. It is anticipated that just such questions about equity and fairness associated with the potential emergence of more than one host site may be raised by the twelve groups of university, nonprofit and private sector partners (Consortia) working to support community engagement, inclusively involve stakeholders, build relationships, and develop innovative forms of mutual learning and public capacity to participate in the consent-based siting process for a federal consolidated interim storage facility during the current Planning & Capacity Building stage of the consent-based siting process. In terms of DOE's Integrated Waste Management System (IWMS), sociopolitical equity concepts are a factor in all aspects of IWMS development and operations and will be particularly applicable should two or more host sites become a possibility, either by the emergence of multiple volunteer hosts or by pursuit as a IWMS program strategy. This paper summarizes recent analyses that explore how four equity metrics - number of states, division of the projected SNF total by the year 2083, current population and land area - might be perceived by host communities, states and regions if two, three or four sites for hosting federal consolidated interim storage facilities were to be contemporaneously realized. Existing institutional arrangements, including Nuclear Regulatory Commission regions, Federal Energy Regulatory Commission regions and Low-Level Radioactive Waste Disposal Compacts, were selected to create hypothetically merged two, three and four-region scenarios to calculate how the four metrics balance out. Projected SNF burden was prioritized with regional contiguity of the consolidated regions a requirement for a functional scenario. The intent was not to promote or suggest any construct as a program objective; the configurations are geopolitical abstractions only to explore potential perceptions of equity. Understanding the possible issues of equity that could arise may benefit program efforts toward achieving a cooperative federalism wherein both the federal and multiple State governments share the goal of manifesting more than one federal consolidated interim storage facility. The main finding is that it is not possible to optimize for all four metrics simultaneously. The analyses demonstrate that trying to create a sense of equity by backfitting a solution to a random population and land distribution will always contain a degree of artificiality.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Long time scale multiphysics simulation of spent nuclear fuel canister in MOOSE

Pebble-bed reactors are an important class of advanced reactors under consideration for various applications where their fuel would give a significant advantage in siting and high-quality heat production. However, the disposal of their fuel is not as thoroughly studied as other fuel forms. In this study, pebble fuel is analysed in a well known spent fuel canister design to characterize the behavior of this fuel form over a long time scale. The results indicate that after approximately 100 years, decay heat is significantly reduced and the maximum temperature in the canister equalizes with the external temperature. The simulation goes on to an end time of a million years, demonstrating the capability of dealing with long time scales efficiently. We conclude that the canister temperatures seem manageable even with very aggressive loading times and while there are several improvements to be implemented in the future, MOOSE is technically capable of simulating the required scenarios.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Spent Fuel Management from Advanced Reactors

Presentation to the IAEA Workshop on the Role of Technological Innovation in Reducing Costs and Improving the Economics of Nuclear Power Generation on spent nuclear fuel management from advanced reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Milestone 1.2.16. Reconciling the Impacts of Thermal Pretreatment on Radiation-Induced H2 Generation from Aluminum-Clad Spent Nuclear Fuel Surrogate Materials

To support the technical basis for the extended dry storage of aluminum-clad spent nuclear fuel (ASNF), thermal pretreatment procedures to minimize the radiation-induced generation of molecular hydrogen (H2) have been investigated. The aim of thermal pretreatment is to eliminate the residual adsorbed water content on the ASNF’s corrosion layers, precursors for H2 generation. To date, irradiation studies in this area have found conflicting results for the effectiveness of thermal pretreatment procedures. The aim of this study was to reconcile those differences. However, the presented results, which utilized a modified in situ thermal pretreatment procedure, afforded H2 yield data that further indicates that thermal pretreatment does not significantly reduce the radiation-induced yield of H2 from gamma irradiated ASNF surrogate materials. Assessment of the differences between thermal pretreatment studies suggests that stainless-steel—present in the irradiation setup of studies that demonstrated a reduction in the yield of H2 with thermal pretreatment—may afford not only unanticipated interfacial chemistry, but also the formation and radiolytic contribution of iron oxides to the chemistry underpinning the formation of H2 in these systems. Given the Department of Energy Standard Canister—proposed for the extended dry storage of ASNF—is predominantly composed of stainless-steel, the potential contribution of stainless-steel and its corrosion layers to radiolytic H2 production should be further investigated. This research was funded by the U.S. Department of Environmental Management, Office of Technology Development, under contract DE-AC07-05ID14517.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Spent nuclear fuel receipt rate analysis within an integrated waste management system (IWMS) architecture that includes consolidated storage

A key parameter in analyzing the performance of an integrated waste management system (IWMS) architecture for the disposition of spent nuclear fuel (SNF) is the SNF receipt rate from reactor and other custodian sites. Receipt rate in this paper means how much SNF is accepted per year for transport in the IWMS from such sites. Introducing one or more federal consolidated interim storage facilities (CISFs) into the IWMS architecture can potentially accelerate the receipt rate profile over time relative to system architectures without a CISF. This raises the question of what an optimal SNF receipt rate profile for an IWMS architecture might be in view of practical constraints and desired system performance attributes and associated metrics. This paper describes a sensitivity study on SNF receipt rates and the associated results for a selected set of IWMS scenarios aimed at informing near-term planning for interim storage capabilities and transportation assets. Two different strategies for CISF operation while awaiting availability of a disposal system to receive SNF are compared: one that relatively quickly fills an initial CISF and then idles the transportation system; and another that aims for more continuous use of transportation assets and receipt capabilities at the CISF. This study examines cost considerations and other factors, such as the timing of clearing reactor sites of SNF, efficient use of capital assets, and some other metrics that might be important to a CISF host community. Based on the analysis, an initial approach is presented that targets a continuous receipt strategy while maintaining the flexibility to step up receipt capabilities to a reasonable degree when needed and beneficial, within overall system constraints.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

In-situ ion irradiation of a spent UO2 fuel: evolution of fission products and nanograins with radiation dose

This study investigates the behavior of fission gas bubbles, five metal precipitates (5MPs) (Mo, Ru, Rh, Tc, Pd) and nanograins in spent uranium dioxide (UO2) fuel under various ion irradiation doses, temperatures, and flux conditions. Utilizing in-situ ion irradiation and advanced transmission electron microscopy, we analyzed the evolution of fission gas bubbles, 5MPs and nanograins in UO2 samples from the Belgium Reactor 3 (BR-3). Our findings reveal significant shrinkage of fission gas bubbles and 5MPs with increasing irradiation dose, accompanied by a decrease in pair density. The study also reveals nanograin coarsening under irradiation. We demonstrate that ion irradiation induces a homogeneous re-solution process where individual atoms are ejected from bubbles and precipitates, leading to their dissolution and subsequent re-precipitation in the matrix. This study provides critical insights into the dynamic behavior of fission products and nanograins under irradiation, facilitating the development of predictive models and contributing to the optimization of nuclear fuel performance and safety.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Spent Nuclear Fuel Receipt Rate Analysis within an Integrated Waste Manage-ment System Architecture that Includes Consolidated Interim Storage

A key parameter in analyzing the performance of an integrated waste management system (IWMS) architecture for the disposition of spent nuclear fuel (SNF) is the SNF receipt rate from reactor and other custodian sites. The introduction of one or more federal consolidated interim storage facilities (CISFs) into the IWMS architecture can enable the receipt rate profile as a function of time to be accelerated relative to system architectures without a CISF. The question then arises as to what an optimal SNF receipt rate profile for an IWMS architecture might be in view of practical constraints and desired system performance attributes and associated metrics. This paper describes a sensitivity study on SNF receipt rates and the associated results for a selected set of IWMS scenarios aimed an in-forming near-term planning for interim storage capabilities and transportation assets. Two different strategies are compared, one that fills an initial CISF quickly and then idles the transportation system while a disposal system is prepared, and a second strategy that aims to provide a more continuous use of transportation assets and receipt capabilities at the IWMS while the disposal system is readied for SNF receipt. Cost considerations and other factors such as impact on timing of clearing reactor sites of SNF, efficient use of capital assets, and other metrics, including those which may be important to a CISF host community, are examined. Based on the analysis, an initial approach is presented targeting a continuous receipt strategy while having the flexibility to step up receipt capabilities to a reasonable degree when needed and beneficial within overall system constraints.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Scalable Direct Recycling of Cathode Black Mass from Spent Lithium‐Ion Batteries

Abstract End of life (EoL) lithium‐ion batteries (LIBs) are piling up at an intimidating rate, which is alarming for environmental health. With further expected rapid growth of LIB use, the magnitude of spent battery accumulation is also expected to grow. LiNi x Co y Mn z O 2 (NCM) cathode materials are a dominant chemistry in high energy LIBs, and make up a huge portion of this waste accumulation. Direct recycling is one of the most promising ways to turn this waste to wealth, but has been limited to lab‐scale, due to lack of robustness, namely the tedious pretreatment required that involves toxic organic solvents. Herein, a process that integrates the pretreatment and relithiation of the cathode black mass is demonstrated. Cathode material from EoL electric vehicle (EV) batteries is treated in a 100 g per batch operation and the regenerated cathode active material demonstrates 100% electrochemical performance recovery, with 91% yield rate, and shows promise for further scale up. This process has the advantages of integration, scalability, and universality, which clears the barricade for direct recycling to move from lab to industry scale with considerable profitability.

Chemistry↗

Zurückgezogen: Glutarimidedioxime: A Complexing and Reducing Reagent for Plutonium Recovery from Spent Nuclear Fuel Reprocessing

Abstract Efficient separation processes for recovering uranium and plutonium from spent nuclear fuel are essential to the development of advanced nuclear fuel cycles. The performance characteristics of a new salt‐free complexing and reducing reagent, glutarimidedioxime (H 2 A), are reported for recovering plutonium in a PUREX process. With a phase ratio of organic to aqueous of up to 10:1, plutonium can be effectively stripped from 30 % tributyl phosphate (TBP) in kerosene into 1 m HNO 3 with H 2 A. The complexation‐reduction mechanism is illustrated with the combination of UV/Vis absorption spectra and the crystal structure of a Pu IV complex with the reagent. The fast stripping rate and the high efficiency for stripping Pu IV , through the complexation‐reduction mechanism, is suitable for use in centrifugal contactors with very short contact/resident times, thereby offering significant advantages over conventional processes.

Xian, Liang↗

Glutarimidedioxime: A Complexing and Reducing Reagent for Plutonium Recovery from Spent Nuclear Fuel Reprocessing

Abstract Efficient separation processes for recovering uranium and plutonium from spent nuclear fuel are essential to the development of advanced nuclear fuel cycles. The performance characteristics of a new salt‐free complexing and reducing reagent, glutarimidedioxime (H 2 A), are reported for recovering plutonium in a PUREX process. With a phase ratio of organic to aqueous of up to 10:1, plutonium can be effectively stripped from 30 % tributyl phosphate (TBP) in kerosene into 1 m HNO 3 with H 2 A. The complexation‐reduction mechanism is illustrated with the combination of UV/Vis absorption spectra and the crystal structure of a Pu IV complex with the reagent. The fast stripping rate and the high efficiency for stripping Pu IV , through the complexation‐reduction mechanism, is suitable for use in centrifugal contactors with very short contact/resident times, thereby offering significant advantages over conventional processes.

Xian, Liang↗

Reciprocal Ternary Molten Salts Enable the Direct Upcycling of Spent Lithium‐Nickel‐Manganese‐Cobalt Oxide (NMC) Mixtures to Make NMC 622

Cathode active material is the most valuable component of spent lithium‐ion batteries, accounting for ≈30% of their overall value. Direct recycling of cathode materials involves recovering, regenerating, and reusing them without breaking down their chemical structure. This approach maximizes the added value of the cathode compound and reduces manufacturing costs by avoiding the need for virgin material production. However, one key challenge in scaling direct recycling from lab to industry is the requirement for highly purified cathode materials, contrasting with the low purity of black mass generated from battery shredding. No efficient separation process currently exists to isolate different lithium‐nickel‐manganese‐cobalt oxides (NMCs) from each other. Thus, direct recycling technologies that can operate with mixtures of multiple NMC stoichiometries will be best‐suited for industrial adoption. This study explores the direct recycling of NMC mixtures into NMC 622 using a “reciprocal ternary molten salts (RTMS)” system. Ionothermal relithiation and upcycling within the RTMS system successfully restore the layered structure, lithium content, and electrochemical performance of degraded NMCs, yielding results comparable to pristine NMC 622 (P‐NMC 622).

25 ENERGY STORAGE↗

Local Gas Purification System in Spent Nitride Fuel Oxidation

A local gas-purification system of the voloxidation operation is examined within the framework of the technology of regeneration of spent mixed uranium-plutonium nitride fuel from fast reactors. The system includes a metal-fabric filter, HEPA filter, ruthenium capture unit, an adsorber for capturing iodine, a unit for additional oxidation and trapping of hydrogen (tritium), and a unit for capturing {sup 14}CO{sub 2}. A description is given of each unit, the overall scheme, and the possibilities of the gas-purification system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Flexible classification with spatial quantile comparison and novel statistical features applied to spent nuclear fuel analysis

Multivariate classification algorithms are a common tool in contemporary data science that have begun to be applied to nuclear forensic analyses. Using a multivariate generalization of quantile–quantile plots for comparing unknown statistical distributions, a new model-free multivariate classifier called the Quantile–Quantile Comparator has been developed and tested on the analysis of simulated irradiated nuclear fuel. A method to estimate the sample-specific probability of having made an incorrect classification decision by analyzing the distribution of quantile comparisons has been developed, and methods have been developed to analyze the partition of the known data library used to analyze unknown data. By analyzing this partition, the analyst may identify classes that are likely to be confused, identify test data that are inconsistent with the data library and return a “None-of-the-Above” classification decision, and optimize the class boundaries. In this paper, the development of these capabilities is described, and a case study using real experimental assay data collected from nuclear fuel extracted from the newly-released In conclusion, Spent Fuel Isotopic Composition Database to demonstrate the efficacy of these capabilities is presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Gamma-ray spectra analyses of molten salts in spent nuclear fuels pyroprocessing facilities for mass measurement

To evaluate radioactive tracer dilution and gamma spectroscopy as a safeguards technique for monitoring the mass of liquid salt in molten salt systems, the gamma spectroscopy data of salt samples from the pyroprocessing facilities in Idaho National Laboratory were acquired and analyzed. The primary focus is the uncertainty analysis of the gamma radioactivity of 154 Eu isotope in the molten salts. The facilities include an electrorefiner for pyroprocessing spent oxide fuels (SOF-ER) and Mark-IV ER for processing the metallic fuels from Experimental Breeder Reactor-II (EBR-II). The relative gamma radioactivity uncertainty was consistently at 3% for 137 Cs for both Mark-IV ER and SOF-ER salts. However, for 154 Eu, it was 8% for SOF-ER salt and 7% for Mark-IV ER salt. Here, the main reason for the higher uncertainty for 154 Eu is believed to be the lower counting statistics in gamma spectroscopy. To improve the gamma radioactivity uncertainty of isotope 154 Eu, a longer data acquisition time (12 h instead of 4 h) and a higher count rate gamma detector for gamma spectroscopy were tested and it was found that the uncertainty for 154 Eu was significantly improved from 7% or 8–3%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Salt mass determination by 22 Na-based radioactive tracer dilution of a highly radioactive molten salt system for pyroprocessing spent oxide nuclear fuels

To demonstrate the feasibility of the 22 Na-based radioactive tracer dilution (RTD) for determining the total salt mass in a 60-kg oxide reduction (OR) system at Idaho National Laboratory (INL) for pyroprocessing spent oxide nuclear fuels, a LiCl- 22 NaCl tracer salt was prepared and spiked to the OR vessel, and salt samples were analyzed by gamma spectroscopy. Due to the high radioactivity (around 10 mCi/g) of the OR salt, mainly from 137 Cs, traditional gamma spectroscopy methods, which typically involve diluted salt samples, was not effective for analyzing the RTD samples. Since the 22 Na tracer radioactivity was very low, undiluted samples were necessary to detect the small amount of 22 Na tracer. Here, despite the high gamma radioactivity of the OR salt, increasing the detector-sample distance from 20 cm to 40 cm during gamma spectroscopy allowed the detection of radioactivity as low as 0.12 µCi/g 22 Na. The preparation of the 22 NaCl-LiCl tracer salt suitable for spiking into the OR salt was successful, and the mass determined by RTD was consistent with that estimated by salt level measurement. Therefore, the 22 Na-based RTD technique for total mass determination of the 60-kg molten salt system for OR is feasible, though more development work, particularly in the improving the uncertainty, is needed .

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Titanium Extraction from Spent Selective Catalytic Reduction Catalysts in a NaOH Molten-Salt System: Thermodynamic, Experimental, and Kinetic Studies

In this study, sodium hydroxide was used as an alkali reagent in roasting spent selective catalytic reduction (SCR) catalysts for titanium extraction. The thermodynamic analysis of the roasting process is discussed in terms of process design. The effects of various roasting parameters, such as the mass ratio of NaOH-to-SCR, temperature, and time, on the extraction efficiency of titanium were investigated; under the optimum parameters of 1.8:1, 550 °C, and 10 minutes, the titanium extraction yield reached 97.8 pct. The roasting kinetics of titanium extraction were also studied. The kinetics of titanium extraction follow the Avrami model, which is expressed as follows: −ln(1−x)=K{sub 0}C{sub m{sub N{sub a{sub O{sub H/m{sub S{sub C{sub R}{sup −2.012}}}}}}}}exp(−(4120/(8.314T)))t, and the overall roasting process is controlled by diffusion in the solid state, with an activation energy (E{sub a}) of 4.12 kJ/mol.

36 MATERIALS SCIENCE↗

Recovery of Carbon and Valuable Components from Spent Pot Lining by Leaching with Acidic Aluminum Anodizing Wastewaters

A three-step process was employed to separate cryolite from used carbon cathodes, known as spent pot lining (SPL), and obtain valuable carbon. The process comprised leaching of NaF from the imbedded electrolyte with water, followed by leaching of Na{sub 3}AlF{sub 6}, CaF{sub 2}, and NaAl{sub 11}O{sub 17} with acidic anodizing wastewater, and then precipitating the electrolyte components from the mixed filtrate from the previous two steps. The influences of stirring rate, liquid–solid ratio, temperature, and time on the extent of leaching of cryolite and recovery of carbon were studied. Additionally, the effects of pH value, F/Al ratio, temperature, and time on the recovery of valuable components in the mixed filtrate were evaluated. The results showed that most NaF in the SPL was dissolved by water leaching. The residual electrolyte in SPL was mainly cryolite and contained approximately 0.95 pct NaF. The purity of the carbon obtained reached 95.5 pct under optimal experimental conditions (leaching temperature: 80 °C; stirring rate: 300 rpm; liquid–solid ratio: 8 mL/g; leaching time: 180 minutes). The recovery of cryolite and the purity of the sodium sulfate crystal from the mixed filtrate were 98.4 and 92.0 pct, respectively, under suitable conditions (pH 9; 75 °C; 4 hour; F/Al ratio of 6:1).

36 MATERIALS SCIENCE↗

Radiation dose rate distributions of spent fuel dry casks estimated with MAVRIC based on detailed geometry and continuous-energy models

This study presents a detailed comparison of the dose rate distributions of a dry TN–32 fuel cask with two geometry models and two cross-sectional datasets. The accuracies of radiation dose rate estimation and computational efficiencies of each geometry model with two cross-sectional data-sets are compared. The use of automated variance reduction techniques can significantly improve the computational efficiency of such a realistic, deep penetration problem that involves radiation transport from different volumetric sources, thereby eliciting only a small statistical error. Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC) is a computational sequence within the SCALE 6.2 code package based on consistent adjoint driven importance sampling (CADIS), a type of automated variance reduction technique. Homogenous and full fuel assembly models are built herein, and two nuclear cross-section libraries (V7–200N47G and continuous energy) are applied in this work. Based on the detailed comparisons, we found that neutron dose rate estimation is more dependent on geometry modeling than on cross-section data. For neutron-induced gamma rays, the dose rate distribution depends on both the spatial self-shielding effect and the cross-section library. The primary gamma rays respectively contribute to the total dose rate by ~91% and ~ 99% on side and top surfaces, and the dose rate accuracy is more dependent on the cross-section library than on geometry modeling. In terms of the computation efficiency and efforts spent on geometry modeling, the homogenous fuel assembly model with the MG library can produce an acceptable dose rate distribution, but the detailed fuel assembly model with the continuous-energy library is required for more precise dose rate estimation.

61 RADIATION PROTECTION AND DOSIMETRY↗