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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
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A new zirconium resonance ionization scheme reveals shifting isotope ratios in spent nuclear fuel towards the fuel pellets edge.
The recovery of lithium from spent lithium-ion batteries (LIBs) is a critical step in advancing sustainability within the battery industry. Traditional lithium extraction methods from end-of-life LIBs predominantly rely on chemical leaching techniques. However, these methods often involve the excessive use of acids, leading to substantial environmental concerns. Additionally, their non-selective nature can compromise the purity of the recovered lithium salt. To achieve battery-grade purity, further purification and recovery processes are necessary. In this study, we introduce a universal and eco-friendly process for lithium recovery, employing terephthalic acid to selectively extract lithium prior to the recycling of other valuable metals. This innovative method achieves lithium recovery rates exceeding 98.53% from layered oxide cathodes and 98.53% from lithium iron phosphate cathodes, delivering an exceptional purity level of 99.95%. By demonstrating applicability across a variety of cathode materials, this approach establishes a universal, sustainable and efficient solution for LIB recycling. The high-purity lithium extraction enabled by this process supports the comprehensive utilization of valuable resources, contributing significantly to the development of a circular economy for battery materials.
This paper summarized the efforts performed to understand decay heat estimation from existing spent nuclear fuel (SNF), under the auspices of the Working Party on Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. Needs for precise estimations are related to safety, cost, and optimization of SNF handling, storage, and repository. The physical origins of decay heat (a more correct denomination would be decay power) are then introduced, to identify its main contributors (fission products and actinides) and time-dependent evolution. Due to limited absolute prediction capabilities, experimental information is crucial; measurement facilities and methods are then presented, highlighting both their relevance and our need for maintaining the unique current full-scale facility and developing new ones. The third part of this report is dedicated to the computational aspect of the decay heat estimation: calculation methods, codes, and validation. Different approaches and implementations currently exist for these three aspects, directly impacting our capabilities to predict decay heat and to inform decision-makers. Finally, recommendations from the expert community are proposed, potentially guiding future experimental and computational developments. One of the most important outcomes of this work is the consensus among participants on the need to reduce biases and uncertainties for the estimated SNF decay heat. If it is agreed that uncertainties (being one standard deviation) are on average small (less than a few percent), they still substantially impact various applications when one needs to consider up to three standard deviations, thus covering more than 95% of cases. The second main finding is the need of new decay heat measurements and validation for cases corresponding to more modern fuel characteristics: higher initial enrichment, higher average burnup, as well as shorter and longer cooling time. Similar needs exist for fuel types without public experimental data, such as MOX, VVER, or CANDU fuels. A third outcome is related to SNF assemblies for which no direct validation can be performed, representing the vast majority of cases (due to the large number of SNF assemblies currently stored, or too short or too long cooling periods of interest). A few solutions are possible, depending on the application. For the final repository, systematic measurements of quantities related to decay heat can be performed, such as neutron or gamma emission. This would provide indications of the SNF decay heat at the time of encapsulation. For other applications (short- or long-term cooling), the community would benefit from applying consistent and accepted recommendations on calculation methods, for both decay heat and uncertainties. This would improve the understanding of the results and make comparisons easier.
The SCALE code system developed at Oak Ridge National Laboratory includes state-of-the-art capabilities for radiation source term and radiation transport simulations that can be used in numerous applications, including dose rate analyses of complex consolidated interim storage facilities (CISFs). A licensed CISF could be used to store tens of thousands of tonnes of spent nuclear fuel discharged from commercial power reactors using various cask and storage pad designs. A CISF design must comply with the regulatory requirements provided in 10 CFR Part 72, including requirements related to annual dose limits applicable to real individuals located beyond the area controlled by the licensee. Therefore, calculating a dose to the public is a necessary part of the licensing process for the construction of a CISF. These calculations are very challenging because of the complexity of the CISF design and the low magnitude of dose rate at large distances from the facility. This paper describes detailed far-field dose rate calculations performed for a proposed CISF using MAVRIC, the Monte Carlo radiation shielding sequence in SCALE 6.2.3, with automated variance reduction based on discrete ordinates calculations. The method presented in this paper uses a detailed Monte Carlo radiation transport simulation in one step from source to dose rate. A series of independent simulations was made using the complete site geometry (all casks present), but with only one cask containing radiation sources to obtain the dose rate maps produced by each storage cask. The CISF dose rate map was obtained by adding the dose rate maps produced by the independent individual cask simulations. Ample volumes of air and soil extending beyond the location of interest for dose rate calculation were included in the calculation model to properly simulate important radiation attenuation and scattering events that affect far-field dose rates. Furthermore, a comprehensive sensitivity study is included in this paper to illustrate the importance of selecting appropriate air volume, mass density, and composition for CISF skyshine dose rate calculations. Dry soil and soil containing water were analyzed to determine their effects on groundshine radiation.
With the forecasted increase in the construction and operation of nuclear reactors, there will be a corresponding increase in the quantity of spent nuclear fuel (SNF) that requires long-term storage. In SNF, transuranic isotopes contribute the most to the long-term radiotoxicity of the fuel and pose a proliferation risk. One option that has been explored to address these issues is the removal of the transuranic isotopes from SNF and the conversion of these isotopes into transuranic fuel (TRU fuel). Here, this work sought to determine how effective a micro-modular Pebble-Bed High-Temperature Gas-Cooled Reactor (PB-HTGR); the 10-MW High Temperature Gas-cooled Test Reactor (HTR-10); and a salt-cooled small-modular pebble-bed reactor (PBR), i.e. the generic Fluoride-cooled High-temperature Reactor (gFHR), are at reducing the inventory of transuranic isotopes while still maintaining the intrinsic safety features of the PBR designs, such as negative temperature coefficients of reactivity. Optimized pebble designs utilizing TRU fuel were found for both reactors through the adjustment for the packing fraction of fuel in each pebble. The Axial Zone Equilibrium Modeling (A-ZEM) method was used in this work to help select the optimized pebble design. Once an optimized pebble design was selected and an equilibrium model was produced, the results from the deep burn (DB) HTR-10 and gFHR designs were compared to the results of two models from the literature. While both the DB gFHR and the DB HTR-10 were able to reduce the weapons-usable transuranic inventory, the performance of these reactors did not match that of the small-modular PB-HTGRs in the literature. Therefore, a need was identified for further refinement of the gFHR design using TRU fuel, as the results for this model were more promising than those of the DB HTR-10, which was strongly limited by the high leakage intrinsic to micro-modular PB-HTGRs.
In 2013, the staff of the U.S. Nuclear Regulatory Commission estimated the reduction of the off-site economic losses from a fire in a drained U.S. spent fuel pool if fuel that had cooled for more than five years were transferred to dry cask storage—an option it called “expedited transfer.” In this article, it is shown that the savings would be much higher than the NRC estimated. Savings increase to about $2 trillion if: losses beyond 50 miles are included; the land-contamination threshold for long-term population relocation is changed to that used for the Chernobyl and Fukushima accidents and recommended by the U.S. Environmental Protection Agency; and, based on the experience of Japan, decontamination of land areas to levels acceptable for population return is assumed to take at least four years. If expedited transfer were implemented, the off-site economic losses would be reduced by about 98%.
X-ray free-electron laser sources such as the Linac Coherent Light Source offer very exciting possibilities for unique research. However, beam time at such facilities is very limited and in high demand. This has led to significant efforts towards beam multiplexing of various forms. One such effort involves re-using the so-called spent beam that passes through the hole in an area detector after a weak interaction with a primary sample. This beam can be refocused into a secondary interaction region and used for a second, independent experiment operating in series. The beam profile of this refocused beam was characterized for a particular experimental geometry at the Coherent X-ray Imaging instrument at LCLS. A demonstration of this multiplexing capability was performed with two simultaneous serial femtosecond crystallography experiments, both yielding interpretable data of sufficient quality to produce electron density maps.
A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. Because heat treatment for stress relief is not required for the construction of the MPC, the canister is susceptible to stress corrosion cracking in the weld or heat affected zone regions under long-term storage conditions. Logic for flaw acceptance is developed should crack-like flaws be detected by Inservice Inspection. The procedure recommended by API 579-1/ASME FFS-1, Fitness-for-Service, is used to calculate the instability crack length or depth by failure assessment diagram. It is demonstrated that the welding residual stress has a strong influence on the results.
In this paper the method of activity measurement of spent radionuclide sources (SRS) based on Co-60 and Cs- 137 placed in well-type storage facilities, in order to forecast the SRS activity which can be placed at this facility is described. The distribution of the absorbed dose through the absorption material is measured with the help of the described absorption spectrometer. This information allows determining the activity of individual radionuclides with the help of mathematical methods. (authors)
The potential for stress corrosion cracking (SCC) of welded stainless-steel interim storage containers for spent nuclear fuel (SNF) has been identified as a high priority data gap. This paper presents a fusion welding process that was developed for SNF canister repair. Submerged arc welding (SAW) was developed to weld 12.7 mm (0.5 in.) thick 304L stainless steel plates to simulate the initial welds on SNF canisters. The SAW procedure was qualified following ASME Boiler and Pressure Vessel Code requirements. During SAW, the welding temperature was recorded at various locations by using thermocouples. After SAW, weld microstructures were characterized, joint mechanical properties were tested, and the maximum tensile residual stress direction was identified. After SAW procedure qualification, artificial cracks were excavated perpendicular to the maximum tensile residual stress direction in the SAW heat affected zone. Machine cold-wire gas tungsten arc welding (CW-GTAW) was developed and used for repair welding at cracked locations.
The U.S. Department of Energy (DOE) is considering the development of one or more federal consolidated interim storage facilities (CISFs) to be used to store commercial spent nuclear fuel (SNF) at locations in the U.S. One of the first technical challenges of a CISF is performing an inspection of SNF canisters upon their receipt to confirm they can be placed into the CISF’s licensed storage configuration. The canister receipt inspection is critical to CISF site operations. The test is conceived as being a helium (He) leak check, intended to confirm that the confinement boundary of a SNF canister is intact. SNF canisters are filled with He when they are sealed, so detection of a He leak indicates that a through-wall flaw has occurred in the canister confinement boundary. Other measurements are planned to occur upon canister receipt in addition to the He leak check such as krypton-85 measurements, which would indicate confinement breaches of one or more fuel rods in addition to a breach of the SNF canister. However, the He leak check has been identified as one of such high importance and has such significant technical challenges that a full-scale demonstration is needed to confirm the He leak test’s viability and to assist in planning relative to its operational requirements. A modeling methodology for simulating the He detection test was developed to help inform the test plan and the design of the test vessels. To develop the modeling methodology a detailed computational fluid dynamics (CFD) benchmark model was constructed to compare against leak rate test data from a transportation package for radioactive material. This report is focused on modeling efforts to simulate the benchmark leak test.
Purpose: To quantify the delivered activity accuracy of Radium 223 dichloride injections, when being administrated for castration – resistant prostate cancer, symptomatic bone metastases. The impact of residual activity in the spent syringe and dispensing accuracy of Ra 223. Methods: The administration is by slow intravenous injection over 1 minute followed by double flushing of the 10 mL syringe and IV with saline. Eighty (80) procedures was used to investigate variations in the activity from the amount prescribed (µCi) = 1.35 × Patient weight Kg. The Activity dispensed into a 10mL syringe using a NIST traceable Capintec CRC-25R Chamber and a cross calibrated capintec CRC-15R to measure activity in the syringe immediately before and after administration Results: The patients weight range from 121Ib to 235lb and doses ranging 74.25 µCi to 144.2 µCi. The deviation of dispensed dose vs Prescribed dose average +2.1% with a range of −1.1% to +5.7%. The Dose measured before administration ranges 79.3 µCi to 154.9 µCi. Deviation from the dispensed dose was show to average +2.9% with a range of −0.8% to +7.3%. The average residual dose post injection was 2.5 µCi or 2.2% of the pre injection activity. Ranging from 0.9 µCi to 6.2 µCi, 0.7% to 5.4% respectively. Subtracting the residual activity from that measured activity before injection and comparing it to prescription dose was shown to have an average variation of +2.7% with a range of −0.8% to 7.4%. Conclusion: The case resulted in the 6.2 µCi maximum residual dose had two syringes. A small, 82.8 µCi activity, case resulted in the 7.4% maximum variation in measures less residual verses prescription dose. The average +2.1 % dispenses activity of Ra 223 over the prescription dosage was seen to counteract the average 2.2% residual dosage found to remain in the syringe.
Spent pickle liquor (SPL) is a waste stream generated in the steel industry that presents both a disposal challenge and a potential resource for metal recovery. This study investigates the electrochemical extraction of high-purity iron metal from SPL using chloride-based molten salt electrolysis (CMSE). Electrochemical characterization of diluted SPL confirmed Fe 2+ as the dominant species. Thermal dehydration of SPL under inert atmosphere yielded FeCl 2 -rich solids, which were directly employed as feedstock for electrolysis in LiCl–KCl eutectic melts at 500 °C. Cyclic voltammetry of FeCl 2 in this melt revealed well-defined Fe 2+ /Fe 0 redox behavior within the electrochemical stability window of the supporting electrolyte. High coulombic efficiency (>85%) electrodeposition was achieved demonstrating that iron metal can be produced from dehydrated SPL by CMSE. The electrodeposited iron exhibited >98 wt% purity, which was further enhanced to 99.9 wt% via arc melting. The resulting iron powder was ferromagnetic, and its size distribution was found to be suitable for powder metallurgy applications. This work demonstrates a scalable, energy-efficient pathway for valorizing SPL into high-purity iron metal, advancing circular economy strategies in the steel industry.
ABSTRACT Secure storage of nuclear spent fuel (NSF) is of great concern for protecting public health and safety. The preferred long-term solution is underground containment in geological repositories, where one or more engineered barrier materials (EBM) encapsulate the NSF and separate it from the natural rock. Bentonite clay is commonly used as an EBM due to its many advantageous properties including low hydraulic conductivity, which ensures limitation of water infiltration to the system and the subsequent risk of corrosion in NSF canisters. However, bentonite clay subjected to heating from nuclear decay may form desiccation cracking. This study conducted disk-shaped free shrinkage tests and ring-shaped restrained shrinkage tests of bentonite clay samples reinforced with basalt microfibers. Digital image correlation was used as a noncontact full-field displacement measurement to track the time-evolving shrinkage and desiccation cracking phenomena and make quantified comparisons between plain bentonite and bentonite with varying contents of basalt microfibers (i.e., 0.0, 0.5, 1.0, and 1.5 % wt.). Results indicate that plain bentonite and basalt microfiber-reinforced samples showed similar free shrinkage behavior, while desiccation cracking behavior was significantly altered by adding basalt microfibers. Microfiber reinforcement effectively reduced major cracks through a “crack-bridging” effect while causing minor cracks to initiate earlier and at higher moisture contents than plain bentonite. Results infer that reinforcing plain bentonite with inorganic microfibers can potentially control desiccation cracking, leading to safer and improved nuclear waste management.
As the title suggests we explored the feasibility of using high-pressure xenon detectors for characterizing spent fuel. The three nondestructive characterization techniques currently in use are the Digital Cerenkov Viewing Device, the Fork Detector and the Safeguards Mox Python (SMOPY) Detector.
Two logistical modeling cases evaluated the relative cost, schedule, and thermal results for strategies that would limit peak salt temperature in a spent-fuel repository to 90°C and 150°C. Runs with the code CALVIN 4.0 show that these low thermal limits are logistically feasible for a salt repository using relatively small waste packages (4-PWR/9-BWR or 12-PWR/24-BWR sizes). Attainment of low-thermal goals was represented using instantaneous waste package thermal power limits at emplacement (3,200 W and 6,800 W) which came from a correlation between package power and peak salt temperature that was developed by a previous study using the finite element method.
A nondestructive assay (NDA) computational study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate dose estimates caused by delayed-neutron (DN) and delayed-gamma (DG) emission for spent nuclear fuel (SNF) that is immersed in nitric acid. The SNF isotopic inventories and delayed-particle (DP) sources are prepared using a generic Westinghouse fuel assembly model for three sets of parameters.
The objective of this work is to improve the accuracy of plutonium assay for nuclear fuel reprocessing. This work addresses such questions as: 1) How much Pu has been recovered in a reprocessing facility?; 2) Is a reprocessing operation consistent with safeguards?; and 3) How much Pu is unaccounted for in a safeguarded facility? This improved Pu accuracy will be achieved by direct measurement of Pu using monochromatic excitation and detection. These elemental measurements will not only provide direct concentration measurements of plutonium in spent fuel dissolver solution. The increased accuracy of the Pu assay addresses the heart of safeguards by providing greater accuracy and less error in the Pu measurements of these facilities. This is further achieved by using the high energy Ka line of Pu at 103.65 keV. This allows for the direct measurement of Pu through container walls such as Zircaloy tubing which is used for the cladding of nuclear fuel rods. While ultrahigh energy X-ray fluorescence (UHEXRF) has been done at synchrotron facilities, X-ray optic technology has advanced to make possible, laboratory based instrumentation to enable such high energy measurements to be done on the bench top. Such instrumentation can change the plutonium assay measurement paradigm by offering safer, more rapid and more accurate Pu measurements through container walls nondestructively.