Journal of Nuclear Fuel Cycle and Waste Technology

Securing the safeguardability of a reprocessing process for spent nuclear fuels (SNFs) is imperative. Particularly, the quantity of special nuclear materials inside SNFs must be estimated with the highest achievable precision. Unlike aqueous reprocessing, pyro-processing involves handling input materials in a solid state. Hence, partially extracted samples analyzed by destructive assay (DA) should maintain an acceptable level of representativeness. In this study, a representative sampling method widely applied in the pharmaceutical industry was adopted for homogenization in the head-end process of pyro-processing. By employing representative sampling, specifically based on the mechanism of the rotary riffler, the overall process of homogenization prior to DA analysis was simplified, and less probable hold-up that could contribute to materials unaccounted for (MUF) would be expected. The resulting Pu sampling uncertainty was confirmed to be less than 1% (for ≥ 1,000 μm particle size and ≤ 5 kg sample mass), ensuring sufficient control of Pu accounting uncertainty at a reasonably low level (≤ 1%). Thus, representative sampling can be a competitive alternative to previously suggested methodologies.

Pyro-processing, Nuclear safeguards, Nuclear material accountancy, Accounting uncertainty, Representative sampling, Homogenization, Heterogeneity

The swelling capacity of bentonite buffers is vital in high-level radioactive waste (HLW) repositories, as it minimizes groundwater infiltration, prevents nuclides from reaching the biosphere, and stabilizes the HLW canisters. As swelling capacity is a function of temperature, understanding bentonite’s behavior at approximately 100°C (its presumed upper limit) is essential. However, research on this subject has been scarce. Hence, this study explored the effects of thermal treatment of Ca-bentonite at 105°C under injected water pressures. The results suggest a 19% reduction in “swell index” and a 35%–36% decrease in the total pressure in thermally treated bentonite. The heated samples demonstrated higher hydraulic conductivity than the non-heated ones, indicating potential performance deterioration in controlling the fluid movement. Furthermore, the injected water pressure (base pressure) was not fully transmitted to the sample owing to the difference between the base and back pressures, leading to variations in the total pressure despite maintaining a constant differential pressure. Thus, the results demonstrated a degradation in bentonite’s swelling capacity and its compromised role in safe HLW disposal, when subjected to treatment at 105°C. The insights from this research can assist in HLW repository design, while highlighting the need for further research into bentonite’s performance.

Thermal treatment, Bentonite, Swelling, Hydraulic conductivity, Injected water pressure

For risk assessment of spent nuclear fuel (SNF) transportation, it is necessary to calculate the damage ratio of SNF rods loaded in the cask. Due to the complexity in the geometry and material properties of SNF, it is impractical to analyze the detailed behavior of every fuel rod and assembly in a single cask model. This study presents a framework for performing cask-level analysis by sequentially simplifying the fuel rods and spent fuel assemblies for fuel damage ratio (FDR) calculation. Using the simplified fuel rod model developed in previous studies, we constructed a CE 16×16 fuel assembly model and presented a methodology to simplify the CE 16×16 assembly model into cuboids. Cask drop analyses were performed to validate the similarity of the detailed CE 16×16 model and the simplified model. Using the proposed simplified models, a procedure for quantifying the bending load and pinch load applied to the fuel rods during the drop impact is presented. The FDR can then be calculated by comparing the quantified loads with their respective failure criteria. Through a case study, the feasibility of the developed framework for systematic and accurate FDR calculation was effectively demonstrated.

Spent nuclear fuel, Risk assessment, Simplified model, Impact acceleration, Fuel damage ratio

Acid-washed solutions containing various ions, including uranium, are produced by washing uranium-bearing soil or minerals with H₂SO₄. Conventional extraction processes are complex, as they involve multiple steps and generate significant amounts of radioactive organic waste, posing environmental risks. Therefore, it is necessary to develop a simplified process that can selectively extract uranium while minimizing radioactive waste production. In this study, thermodynamic equilibrium calculations were performed to investigate the selective extraction of uranium through its reaction with H₂O₂. A basic additive was introduced to facilitate this reaction. The calculation results indicated that uranium in acid-leached solutions could be selectively extracted through a single-step precipitation process, which was experimentally validated. These findings can be utilized to design an efficient process for obtaining high-purity uranium from uranium-bearing soil or minerals.

Acid-washed solution, Uranium-contained soil or mineral, Precipitation, Selective extraction, High-purity uranium

Uranium (U), an essential source for nuclear energy production, poses serious environmental and radiological threat due to its high mobility and long half-life. Uranophane [Ca(UO₂)₂SiO₃(OH)₂·5H₂O], a secondary U silicate mineral, is known as a solubility-limiting phase in U mining deposits and nuclear waste repositories (controlling U immobilization). However, research on uranophane dissolution, particularly under the influence of organic and inorganic ligands, remains lacking. This study investigates uranophane synthesis and its dissolution at pH 8 through batch experiments using organic ligands (citric acid (CA) and humic acid (HA) at 50–150 ppm) and inorganic ligands (carbonate, nitrate, chloride, and silicate at 10−⁴ M to 10−⁶ M). Obtained results suggested that CA and carbonate significantly enhanced U release, reaching 27.6 ppm. Mixed systems containing both organic (50–150 ppm CA) and inorganic (10−⁴ M carbonate) ligands revealed increased U release, however were less effective than single-ligand systems due to competitive interactions with carbonate dominating U speciation. Visual MINTEQ modeling was used to identify uranyl complex species in the solutions. Dissolution rate and kinetic modeling were determined to predict U release trends. These findings emphasize the role of various ligand types in nature and their impact on U mobility, aiding remediation strategies for contaminated environments.

Uranium, Uranyl silicate minerals, Uranophane, Inorganic ligands, Organic ligands, Dissolution

The efficient fabrication of uranium-based liquid fuels and the structural integrity of reactor materials are critical challenges for the deployment of chloride-based molten salt reactors (MSRs). As part of KAERI’s ongoing MSR development, this study investigates an optimized uranium chlorination process and a corrosion assessment of candidate structural materials under conditions more closely resembling actual reactor cores. To enhance process efficiency and scalability, metallic uranium was converted into uranium trihydride (UH₃) via hydriding, achieving 34.1% efficiency. UH₃ was chlorinated with ammonium chloride (NH₄Cl), yielding uranium trichloride (UCl3) with a conversion rate over 98% and purity above 99%, as confirmed by ICP-OES. The UCl₃ was used to fabricate various uranium-based liquid fuels for MSR applications. Simultaneously, the corrosion behavior of SS304, SS316, and Hastelloy-N was evaluated using a natural convection loop filled with a NaCl– MgCl₂ eutectic salt mixture. The system operated for 500 hours at 500–580°C to replicate MSR conditions. Corrosion analysis revealed that SS304 suffered severe degradation, SS316 showed moderate resistance, and Hastelloy-N demonstrated superior stability, although some cold leg samples experienced mass gain due to corrosion product deposition. These findings provide key insights into optimizing liquid fuel synthesis and selecting corrosion-resistant materials for safe, long-term MSR operation.

Molten salt reactor (MSR), Liquid fuel fabrication, Natural convention loop, Corrosion

Uranium-contaminated soil can be cleaned using an acid washing process. However, high-concentration acid washing generates substantial amounts of radioactive waste, making it essential to develop a treatment process using low-concentration acid. This study evaluated the effectiveness of low-concentration sulfuric acid washing for uranium removal from contaminated soil. Experiments were conducted with a 0.05 M sulfuric acid solution. With a mixing ratio of soil to acid solution at 1:5, three consecutive washes were sufficient to remove uranium from contaminated soil to clearance level. During the acid washing process, real-time pH monitoring was performed to analyze the correlation between uranium leaching and pH changes. This led to the establishment of a monitoring-based process control strategy. In conclusion, we identified an effective method for removing uranium from soil under low acid concentration conditions. Consequently, significant reductions in radioactive waste generation are anticipated.

Uranium-contaminated soil, Low-concentration acid washing, Removal, Clearance, pH monitoring

This study used an in-house beta version code, developed on the Microsoft^® Excel platform and based on the Regulatory Guide 1.109 model, for radiation dose calculations. The results were compared with those of NRCDose3 Code Version 1.1.4. Although most results were compatible, five significant discrepancies were identified. First, potential errors in the effective dose for ³ H inhalation and ingestion were due to inadequate incorporation of dose coefficients based on chemical forms or absorption types in the GASPAR module. Second, potential errors in ¹⁴C effective doses resulted from incorrect application of age-specific consumption values and dose coefficients. Third, potential errors for ¹³¹I inhalation doses occurred due to inadequate consideration of dose coefficients for chemical form or absorption type in the GASPAR. Fourth, potential errors in equivalent dose for radionuclides (e.g., ⁶⁰Co and ¹³¹I) were caused by inconsistencies in the ordering of organs or tissues in dose coefficients and output files. Fifth, in the LADTAP module, when “Salt water” was selected and the International Commission on Radiological Protection Publication 72 was applied, liquid effluent doses were incorrectly output for only three age groups instead of six. This study analyzes these errors and proposes interim corrective measures to ensure accuracy pending software revisions.

NRCDose3, Potential error, Root cause analysis, Offsite dose calculation, Radioactive effluent

Uranium-contaminated soil can be effectively decontaminated through acid leaching; however, this process process generates significant amounts of radioactive wastewater. Therefore, developing efficient methods to remove uranium from wastewater is essential to minimize radioactive waste generation. This study investigates the applicability of various precipitation methods for uranium removal from acidic wastewater produced during soil-washing processes. Three methods were evaluated: metal hydroxide (M–OH_x) co-precipitation, uranium peroxide (UO₄) precipitation, and uranium phosphate (KUO₂PO₄) precipitation. The M–OH_x precipitation method removes uranium by precipitating excess metal ions in wastewater by adjusting the pH. This method is easy to use and has a high removal efficiency. The UO₄ and KUO₂PO₄ precipitation methods involve adding reagents to precipitate uranium in the mineral phase. They enable selective uranium separation and further volume reduction. In the results, M–OH_x and KUO₂PO₄ precipitation methods remove uranium to less than 1 mg∙L⁻¹ within 2 h, demonstrating superior capabilities compared to UO₄ precipitation. The optimal method is different depended on the management strategy for the recovered uranium. The M–OH_x precipitation method was suitable for permanent disposal, whereas KUO₂PO₄ could be recycled. Based on these findings, guidelines for the effective treatment of wastewater containing uranium from the soil-washing process can be established.

Soil washing, Wastewater treatment, Uranium precipitation, Studtite, Meta-ankoleite, Metal hydroxide co-precipitation

The corrosivity of molten salt presents a major challenge for the commercialization of molten salt reactors, which utilize molten salt as both fuel and coolant. To protect structural materials of molten salt reactors, minimizing moisture—the primary source of corrosion—is crucial, necessitating precise moisture concentration measurements. This study examines the role of an inert gas atmosphere in analyzing moisture in molten chloride salts. Four chloride salts with different hygroscopic properties (NaCl, KCl, MgCl₂ and ZnCl₂) were tested. Each was analyzed in three states: as-received and dried by heating for 6 and 12 hours. Karl Fischer titration was employed to measure the moisture concentrations in salts under both air and an argon-filled glove box. Results showed consistently lower and more stable moisture concentrations in the inert atmosphere, highlighting the necessity of an argon environment for accurate moisture analysis in molten salts.

Molten salt, Dehydration, Moisture concentration, Karl Fischer titration, Inert gas atmosphere