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

Maintaining the structural integrity of spent nuclear fuel (SNF) cladding under transport and handling conditions is essential for safe and economic management of SNF. Reoriented hydrides significantly reduce the fracture resistance of cladding under pinch loads, necessitating an investigation of its fracture behavior. In this study, a simulation model was developed for cladding under ring compression test (RCT) using ductile damage model. This model enables the simulation of crack initiation and propagation under various stress triaxiality conditions. However, the calibration of the fracture parameter of the developed model is challenging due to the lack of experimental data and complexity of the parameter space. To address this, a metamodel-based optimization framework was proposed to calibrate the fracture parameters of the Zr/hydride interface, which exhibits the lowest load resistance, using RCT data. Two optimization algorithms - global search algorithm (GSA) and a genetic algorithm (GA) - were employed and their execution time, accuracy, and practicality were compared. Both algorithms produced nearly identical solutions. The optimized parameters were validated against experimental RCT data and demonstrated high prediction accuracy for crack initiation load and displacement during RCT.

Spent nuclear fuel, Hydride reorientation, Pinch load, Finite element analysis, Metamodeling, Parameter optimization

For a fundamental assessment of spent nuclear fuel integrity, the fatigue behavior of cladding tubes was investigated by evaluating the fatigue strength reduction factor (K_f) due to wear and oxidation-induced degradation. In this study, as wear geometry significantly influences fatigue life cycles leading to fracture, it was determined based on the contact conditions observed in commercial fuel designs. Wear and oxidation-affected cladding tubes were fabricated, along with reference specimens without wear, for comparative analysis. Fatigue tests were conducted by applying and releasing cyclic high internal rod pressures (<102 MPa) at 4 Hz to simulate operational conditions. The obtained average K_f for wear specimens was 1.04, based on an effective stress analysis relative to the unworn reference tube. However, oxidation led to a significantly greater reduction in fatigue life with K_f values exceeding 1.5, indicating more dominant role in fatigue degradation. It is noteworthy that oxidation‑induced metal wastage caused a significantly greater reduction in fatigue life than wear, thereby emphasizing the need for stringent corrosion control in dry storage environment.

Fatigue, Wear, Zircaloy cladding tube, Oxidation tube

Two-dimensional MXenes (Ti₃C₂T_x) have received extensive attention due to their easy availability, hydrophilic behavior, and tunable chemical properties; however, their permeability-selectivity trade-off and self-stacking property limits their application in water treatment. In this study, three-dimensional Ti₃C₂T_x films with high adsorption capacity, permeability, and high selectivity were successfully fabricated using a simple vacuum-assisted filtration method for the removal of radionuclides (Cs⁺, Sr²⁺, and Co²⁺) from aqueous solutions. Various characterization results suggested that radionuclides tended to coordinate with heterogeneity of the sites (containing hydroxyl, fluorine, and oxygen groups) on Ti₃C₂T_x, as well as the multilayer adsorption properties of the Ti₃C₂T_x films. The synthesized Ti₃C₂T_x film demonstrated maximum removal efficiencies of 99.50%, 99.54%, and 99.8% for Cs⁺, Sr²⁺ and Co²⁺, respectively, at pH 7 and 25℃. The adsorption behavior and process analysis showed that the adsorption mechanism primarily involves electrostatic and chemical interactions on the Ti₃C₂T_x film. Furthermore, the films not only sequestrated radionuclides from the aqueous solution, but also selectively separated target radionuclides by adjusting the Ti₃C₂T_x interlayer size. Therefore, the newly developed Ti₃C₂T_x films demonstrated great potential for the removal of radionuclides from aqueous solutions.

MXene (Ti₃C₂T_x) films, Radionuclides, Adsorption, Separation process, Interlayer

This study investigated the unconfined compression behavior of unsaturated compacted Ca-bentonite (dry density = 1.6 Mg·m⁻³) after exposure to 0.02 M Ca(OH)₂ solution and elevated temperatures (25℃ and 150℃) for up to 28 days. Unsaturated Ca-bentonite demonstrated brittle failure behavior regardless of the experimental conditions. At 25℃, the unconfined compression strength (q_u) of Ca-bentonite specimens mixed with either deionized (DI) water (pH 6) or Ca(OH)₂ solution (pH 12) was statistically unchanged. However, at 150℃, qu values increased significantly compared to 25℃. A similar trend was observed for Young’s modulus, indicating increased stiffness at higher temperatures. The enhancement is attributed primarily to reduced water content and corresponding increases in suction rather than direct mineralogical alteration, as confirmed by X-ray diffraction analysis. This study provides the early-stage mechanical response of unsaturated Ca-bentonite under repository-relevant thermal and alkaline conditions, offering understanding into its performance before saturation is achieved.

Ca–bentonite, High-level radioactive waste disposal, Buffer material, Alkaline solution, Unconfined compressive test, Stress–strain behavior

A method for predicting the level of radioactive waste during the preliminary decommissioning stage of a nuclear reactor is presented, utilizing both Monte Carlo and deterministic codes simultaneously. An autonomous simulation code, developed in Python, was created to estimate radioactive waste during the decommissioning of both PWR and CANDU reactors. The materials considered for activation analysis include only concrete, stainless steel, and vessel steel. Neutron flux is calculated using MCNP, a Monte Carlo-based code, while SCALE-ORIGEN, which is based on the matrix exponential method, is used to perform the activation analysis. The developed RSAM (Reactor Structures Activation Module) serves as a bridge between these two codes, enabling the automatic generation of activation analysis input files and the interpretation of results. Additionally, RSAM can autonomously conduct activation sensitivity analysis based on material data from NUREG-3474. To verify RSAM’s performance, activation analyses were conducted on the structural components of CANDU and OPR-type reactors. The developed Module effectiveness was demonstrated through case studies on CANDU and OPR reactor (→ PWR and PHWR), indicating its broad applicability and potential as a valuable tool for nuclear power plant decommissioning projects.

MCNP, ORIGEN-S, Activation analysis, Impurity, Structure material

Spent ion exchange resins (SIERs) are commonly immobilized in cement, necessitating the reliable determination of radionuclides in cement-based waste forms. This study developed a simple and practical method for quantifying ¹²⁹I, a key radionuclide in low- and intermediate-level radioactive waste (LILW) disposal. Investigation into iodine behavior during leaching, extraction, and precipitation revealed that iodate (IO₃^⁻) readily adsorbs onto cement minerals, reducing chemical recovery in extraction-based methods. To mitigate this, we implemented an in-situ conversion of IO₃^⁻ to I₂, improving chemical yields from 58% to 93% in fine cement samples (φ < 0.5 mm). The optimized method was further applied to cement containing anion exchange resins, where increasing resin content led to a decrease in chemical yield, reaching 51% at 10wt% resin loading, close to the practical limit in ordinary Portland cement (OPC) waste forms. This decline was likely attributed to the re-adsorption of iodide (I^⁻) onto the resins during the conversion of IO₃^⁻ to I₂. Despite this, measured ¹²⁹I activities closely matched expected values, demonstrating its reliability for routine analysis of cement-based radioactive waste.

Characterization of radioactive waste, Radiochemistry, Decommissioning, Silver iodide, Ettringite

With an emphasis on climate change and long-term perspectives, this study discusses the impact of rainfall patterns on the safety and integrity of near-surface radioactive waste disposal facilities in South Korea. Future rainfall and infiltration scenarios are prepared by using historical rainfall data from the Ulsan district, and the impact of rainfall patterns is modeled with COMSOL Multiphysics and GoldSim. Infiltration patterns do not significantly affect the total annual dose. However, they do have a minor impact on the total annual dose from ¹²⁹I and ⁹⁰Sr, contributing a smaller dose during the 1,000-year simulation period. Radionuclides such as ³H, ⁹⁹Tc, and ¹⁴C are the primary contributors to the total annual dose, which results from the radionuclide concentrations in the saturated rock zone, regardless of the assumed cases of inputs such as waste zone permeability and saturated zone groundwater flow rate. This study provides essential insights and recommendations for the safe management and design of multi-layered cover system in radioactive waste disposal facilities, considering the evolution of long-term climatic conditions.

Climate change, Precipitation, Near-surface disposal, Multi-layered cover, Water flow, Radionuclide transport

This study introduces a strain limit-based evaluation method that considers stress triaxiality to assess the structural integrity of radioactive waste transport containers. Unlike conventional stress-based approaches, this method provides a more precise assessment of localized plastic deformation and damage. By incorporating strain-based failure criteria, it enables a realistic evaluation of impact-induced deformation under complex stress states. A post-processing program is developed to intuitively verify strain safety margins, improving assessment accuracy. Additionally, sub-modeling techniques refined the strain distribution analysis in damage-concentrated areas, which could not be captured in full-scale simulations. This enhances the reliability of impact evaluations. Based on these procedures, a drop safety verification process was established for the freight container used in low- and intermediate-level waste transport. The results demonstrated the feasibility of strain limit-based evaluation in practical applications. This approach improves the structural assessment of transport containers and contributes to the advancement of design and certification processes.

Transport container, Radioactive waste, Drop test, Stress triaxiality, Strain limit

The sorption of Europium(III) (Eu(III)) on MX-80 bentonite and granite was systematically studied as a function of ionic strength (0.05–1 mol·kgw⁻¹) and pHm (4–9) in Ca-Na-Cl solutions by batch experiments and 2-site protolysis no electrostatics surface complexation and cation exchange (2SPNE/CE) model. Eu(III) has often been used as a chemical analogue of plutonium(III) which is an important element of interest for the safety assessment of a deep geological repository (DGR) for nuclear waste. It was found that the sorption of Eu(III) on MX-80 and granite was dependent on pH_m. The sorption of Eu(III) on MX-80 was dependent on ionic strength at pH_m ≤ 6. The effect of ionic strength on the sorption of Eu(III) on granite was negligible. The 2SPNE/CE model simulation suggested that the ion exchange reaction might be the main sorption mechanism at low pH_m, while the surface complexation reactions might be the dominant sorption mechanism at higher pHm for the sorption of Eu(III) on both MX-80 and granite.

Europium, Sorption, MX-80, Granite, 2SPNE/CE

Deep geological disposal facilities require early regulatory involvement to ensure safety, transparency, and efficiency throughout their long development process. Leading countries, such as Finland, Sweden, France, and the UK, actively engage regulatory bodies from the early stages, following international guidelines set by the IAEA and OECD/NEA. However, in Korea, the current regulatory framework lacks a legal basis for pre-licensing reviews, highlighting the need for institutional reforms and clear agreements between regulators and operators. A phased regulatory approach which covers pre-site selection, site selection, and post-selection stages can enhance project stability and minimize risks. To successfully implement early regulatory involvement, Korea must revise its legal framework, develop strategic roadmaps, and foster transparent communication with stakeholders.

Deep geological disposal, Regulatory framework, Pre-licensing, Long-term safety

This study integrates MCNPX and MATLAB to analyze decay heat and radioactivity evolution in spent nuclear fuel (SNF) over 10¹⁰ seconds. MCNPX models burnup-dependent isotopic inventories, while MATLAB solves Bateman equations for decay analysis. Six phases, defined by dominant isotopes, dictate safety protocols. For decay heat, the short-term phase (¹⁴⁰La, ²³⁹Np, ¹⁴⁴Pr) requires active cooling; intermediate (¹⁴⁴Pr, ¹⁰⁶Rh, ⁹⁵Nb, ⁹⁵Zr) permits passive dry cask storage; and long-term (¹³⁷Cs, ⁹⁰Sr, ²⁴¹Am, ²⁴¹Pu) necessitates geological repositories. Radioactivity evolves through three phases: immediate with high gamma emitters (⁹⁵Nb, ¹⁴⁴Ce, ¹⁴⁴Pr, ⁹⁵Zr) demanding shielding; intermediate dominated by ¹³⁷Cs, ²⁴¹Pu, ⁹⁰Sr, risking cask integrity; and long-term radiotoxicity from ¹³⁷Cs, ⁹⁰Sr, ²⁴¹Am, ²⁴¹Pu, requiring geological confinement. These temporal transitions highlight phase-specific risks: short-term strategies prioritize active cooling and shielding, while long-term solutions focus on multi-barrier isolation. The study provides a roadmap for optimizing cooling systems, shielding designs, and repository architectures to mitigate risks across the SNF lifecycle. By mapping isotopic contributions over time, the framework supports balanced technical-regulatory decisions, enhancing safety from reactor discharge to permanent disposal.

Spent nuclear fuel, Decay heat, Fuel composition, Spent fuel radioactivity

In preparation for nuclear power plant decommissioning, an accurate assessment of radionuclide inventories in reactor components is essential. This study investigates the spatial distribution of activated radionuclides in the bioshield concrete of a pressurized water reactor (PWR). Using MCNP6.3 and ORIGEN in SCALE 6.3.1, neutron flux and radionuclide inventories are calculated over 7,500 mesh elements within the concrete region. A total of 35 radionuclide inventory maps are produced, providing detailed visualization of radionuclide distributions. Among the analyzed radionuclides, ³H, ⁴¹Ca, ⁶⁰Co, ¹⁵²Eu, ¹⁵⁴Eu, and ¹³⁴Cs exhibit the highest specific activity levels. The analysis also performs sensitivity evaluations of the radionuclide inventory in the bioshield concrete concerning impurity content, neutron irradiation time, and cooling period, and it presents the resulting radial distribution characteristics of radionuclides in the reactor. In addition, a comparison between ORIGEN versions 6.1 and 6.3.1 showed differences in inventory estimations, underscoring the importance of using updated nuclear data. The results are expected to offer essential data for radioactive concrete waste classification and disposal planning.

Radionuclide inventory, Specific activity distribution map, Bioshield concrete, decommissioning, MCNP, ORIGEN

This study evaluates calibration models for the triple-to-double coincidence ratio (TDCR) method in liquid scintillation counting, focusing on the relationship between TDCR values and detection efficiency for ³H and ¹⁴C. Four datasets were analyzed: three reconstructed from published literature and one obtained through original experimental measurements. A nonlinear core function (CF) model was compared with first- to third-order polynomial regressions. Model performance was assessed using root mean squared error (RMSE), mean absolute error (MAE), symmetric mean absolute percentage error (SMAPE), and the Akaike and Bayesian information criteria (AIC and BIC). While the CF model showed superior performance for ³H due to its inherently nonlinear response, simpler polynomial models—particularly linear and quadratic—yielded comparable accuracy for ¹⁴C across all datasets. These models also enable analytical uncertainty propagation and offer greater numerical stability. The findings support a model selection strategy that emphasizes simplicity and parameter parsimony without sacrificing accuracy. This work highlights the practical advantage of selecting the least complex model that sufficiently captures the efficiency–TDCR relationship.

Liquid scintillation counting, Triple-to-double coincidence ratio, Effieiciency model, Calibration

Liquid scintillation counting (LSC) efficiency for tritium (³H) is calibrated using polynomial quench correction models and uncertainty analysis. We employed a previously published dataset of ³H quench standards, covering a wide range of quench levels. Second-, third-, and fourth-order polynomial models were fitted to the calibration data based on the published relationship between the instrument’s quench index and ³H counting efficiency. This was intended to provide representative cases for uncertainty analysis, rather than to identify an optimal functional form. The propagated uncertainties were quantified across the quench range, revealing that the intermediate and high quench regions are particularly sensitive; small errors in quench or calibration yield disproportionately large efficiency uncertainties. In contrast, the low-quenching region exhibited a relatively minor uncertainty contribution. These results highlight the methodological importance of applying rigorous uncertainty propagation to ³H LSC efficiency calibrations. In particular, explicitly accounting for calibration-fit uncertainty via the law of propagation of variances ensures a more reliable activity estimation and alignment with modern metrological standards, which is crucial for the confident quantification of low-level ³H.

Liquid scintillation counting (LSC), Tritium, Quench curve, Uncertainty, Transformed spectral index of the external standard (tSIE)

Boron concentrates are one of the common radioactive waste produced in pressurized water reactor (PWR) NPPs. Liquid waste is generated during the operation of NPPs, including in the normal process, fuel reloading, and maintenance. The liquid waste is composed of different concentrations of boric acids dissolved in the primary coolant as a result of core reactivity control. At the early stage of NPP operation, the boron concentrates are solidified using a cement material. Cement solidification offers high product stability and low cost and is considered a promising treatment technology for very low, low, and intermediate level boron concentrate radioactive waste. However, the waste loading in cement solidification of radioactive waste is usually lowered to obtain high quality cement solidified forms. In this study, a cement re-solidification process for historic cement solidified boron concentrates was systematically investigated. The composition of the cement solidified form was evaluated to obtain a suitable material composition with reasonable cost. A water immersion test and a thermal cycle stability test were performed to determine whether the solidified form is suitable for disposal. The ANS 16.1 test also was implemented to evaluate its chemical stability. It is reasonable to conclude that the fabricated solidified forms have sufficient potential to be applied for the re-solidification process for historic cement solidified boron concentrates.

Cement solidified boron concentrates, Cement, Solidification, ANS 16.1, Disposal

As the storage capacity of spent nuclear fuel pools at nuclear power plants approaches saturation, dry storage systems are gaining increasing attention as an interim solution. In particular, metallic canister-based systems have been widely adopted due to their operational efficiency and design flexibility. However, the global trend toward higher burnup fuels has emphasized the importance of managing the long-term integrity of internal structural components, which are subject to elevated temperatures and restricted inspection access. This study presents the current status of evaluation system development aimed at managing degradation in key structural materials—namely, 17-4 precipitation-hardened stainless steel and 6061-T651 aluminum alloy—used in dry storage systems. Building upon the framework outlined in the U.S. NRC’s NUREG-2214, aging management reviews have been performed to identify critical degradation mechanisms such as thermal aging and creep under both normal and off-normal conditions. To experimentally support these reviews, thermal aging and creep test systems have been constructed and validated. Testing is being carried out under simulated dry and wet-steam environments, with preliminary data being collected on mechanical property degradation and time-dependent creep behavior. Ongoing efforts include system reliability enhancement, oxide characterization and the refinement of technical bases for future application to aging management programs and time-limited aging analyses. These efforts are expected to support the establishment of a domestic AMP and TLAA for dry storage systems, and contribute to the design advancement and infrastructure development of future interim storage facilities in Korea.

Dry storage system, Aging management review (AMR), Thermal aging, Creep, Structural materials degradation

This study investigates the design and thermal optimization of a microwave reactor crucible system developed for treating radioactive carbon-bearing spent activated carbon. To enhance thermal efficiency and uniformity, six technical improvements were examined: optimized insulation placement, integration of internal heating elements, installation of a stainless steel central reflector, partial replacement of quartz with SUS 304, operation under vacuum (100–300 Torr), and crucible rotation. Each parameter was systematically tested under controlled experimental conditions, with performance evaluated by heating rate, power consumption, and thermal uniformity. The results showed that insulation on the crucible's outer wall provided the highest energy retention, while the central reflector most effectively improved heating uniformity. Partial use of SUS 304 reduced crucible manufacturing costs by over 50% and enhanced mechanical durability. Vacuum conditions marginally suppressed convective heat loss, and crucible rotation minimized local overheating. The integrated strategy yielded a 10–30% improvement in thermal efficiency, demonstrating its practical value in high-temperature microwave applications. This study proposes a scalable framework that may be applied to a wide range of high-temperature microwave systems, especially in the field of radioactive waste treatment.

Radioactive spent activated carbon, Microwave heating, Crucible design optimization, Insulation, Reflector, Heating element

The operation and decommissioning of nuclear facilities generates uranium-containing waste. It is important to note that significant quantities of uranium-contaminated soil may be generated from uranium-contaminated ground surfaces. Furthermore, uranium deposits may also be generated from the decontamination of uranium-contaminated soil. These particulate radioactive waste can be disposed of through a non-dispersive immobilization process, such as cementation. In this study, cement-solids were prepared under different mixing ratios for uranium-contaminated soil and uranium deposits. The cement solids were evaluated for workability, free water, and stability. The cement solidification of uranium-contaminated soil achieved a maximum waste loading of 30 wt% with a compressive strength of 16.607 MPa. Uranium precipitates containing about 40 wt% diatomite were immobilized in a cement matrix at a maximum loading of 20 wt%, achieving a compressive strength of 3.83 MPa. Cement solidification of uranium precipitates, mainly composed of U and Fe, achieved a maximum waste loading of 30 wt% with a compressive strength of 7.12 MPa. The results of this study are expected to support the selection of conditions for the solidification of uranium waste.

Uranium-contaminated soil, Cementation, Cemented waste form, Solidification, Precipitate, Disposal, Clearance

In Korea, all spent fuel from pressurized water reactors (PWRs) is stored exclusively in spent fuel pools (SFPs). To expand storage capacity and sustain plant operation, high-density storage racks are being deployed. Ensuring the safety of spent-fuel handling and storage entails establishing hypothetical accident scenarios and performing integrated evaluations of criticality, shielding, thermal performance, and structural behavior. This study evaluates the structural integrity of existing storage racks under seismic impact loads representative of current design-basis conditions. Specifically, the impact load predicted for newly installed racks was conservatively applied to the existing racks, and the resulting structural responses were assessed. The results show that the existing racks—owing to their thicker cells relative to the new racks—maintain sufficient structural integrity under the applied seismic impact load. In addition, a parametric assessment was conducted by varying the impact-load level and reducing rack-cell thickness to examine sensitivity and margin. At the SSE level, the buckling allowable remained satisfied for an ~9% increase in impact load and for an ~8% reduction in rack-cell thickness, indicating potential buckling only beyond those thresholds. These findings support safe SFP operation with high-density configurations.

Axial load, Buckling load, Compression load, Spent fuel, Storage rack, Structural safety

Copper plates are a proposed engineering solution to enhance thermal dissipation from nuclear waste canisters through the surrounding bentonite buffer. However, their long-term effect on the hydraulic integrity of this bentonite buffer is uncertain. This study experimentally investigates the impact of penetrating copper components on the hydraulic conductivity of compact bentonite. We measured the hydraulic conductivity of bentonite blocks penetrated by 0 (control), 5, and 10 copper pins, both at initial saturation and after a 293-day aging period. Initially, hydraulic conductivity decreased from 1.49 × 10⁻¹³ m/s in the control as the number of pins increased. However, this monotonic trend did not persist; after 293 days, the relationship became non-linear, with the 5-pin block showing the lowest conductivity and the 10-pin block the highest. Crucially, our findings provide no evidence that copper penetration or subsequent corrosion systematically increases hydraulic conductivity. We conclude that the integration of copper plates for thermal management is unlikely to compromise the critical containment function of the bentonite barrier.

Engineered barrier system, Hydraulic conductivity, Heat dissipation, Copper plate, Bentonite buffer

This study investigates the potential impact of adopting ICRP Publication 103-based dosimetric data on radiological evaluations for transport casks containing radioactive materials and spent nuclear fuel. Current assessments rely on ICRP Publication 60 and its associated conversion coefficients and regulatory limits; however, ICRP Publication 103 introduces revised radiation weighting factors, updated decay data, and more realistic computational phantoms. To evaluate the implications of this transition, dose rates were calculated using both ICRP 74 and ICRP 116 conversion coefficients for representative gamma- and neutron-emitting sources, as well as for spent nuclear fuel. Additionally, the effects of updated A1 and A2 values based on ICRP Publication 103 were analyzed. The results indicate that changes in dose conversion coefficients yield negligible differences in gamma dose rates and minor variations in neutron dose rates due to compensating spectral effects. Revisions to A1 and A2 values resulted in changes of less than 10%, with minimal influence on containment evaluations for spent nuclear fuel transport casks. Overall, transitioning to the ICRP 103 framework is expected to have little impact on radiological assessment of transport cask for radioactive materials and spent nuclear fuel.

ICRP Publication 103, Transport casks, Dose conversion coefficients, Radiation shielding, Spent nuclear fuel, A values

This paper reviews current regulations governing the transport of spent fuel within nuclear power plant (NPP) sites and proposes measures to improve efficiency without compromising safety. In South Korea, historical practice has required the use of full transport casks even for short on-site transport. While a recent amendment by the Nuclear Safety and Security Commission (NSSC) now permits the use of transfer casks within an NPP site, dose standards and cross-references inherited from off-site transport rules remain overly conservative and internally inconsistent. We propose three actions: (1) define off-site transport at the licensed NPP site boundary rather than at the unit boundary; (2) for on-site transport using transfer casks, apply ALARA-based dose control through the facility’s radiation protection program instead of prescriptive off-site dose tables; and (3) revise the exception clause in Article 9(2) of the technical regulations to remove conflicts with special-measures approvals and to restore internal coherence. Process mapping indicates that these changes will streamline procedures, reduce handling steps and work hours, and lower occupational dose, while preserving safety margins. Clarifying the regulatory scope and aligning dose management with on-site conditions are expected to enhance safe, reliable NPP operation and provide a practicable pathway for optimizing on-site spent fuel transfer operations in Korea.

Spent fuel, Safety regulation, Transfer cask, Transport cask, 10 CFR 71, and SSR-06 (Rev.1)

This study analyzed the trends in radioactive effluent releases and the dose contributions of individual radionuclides from five nuclear power plant sites in Korea over the past five years (2020–2024). Based on the data published in the annual “Environmental Radiation Monitoring Reports” by Korea Hydro & Nuclear Power (KHNP), the cumulative release amounts, emission frequencies, and dose contributions by radionuclide were quantitatively analyzed for both liquid and gaseous releases. The analysis found that tritium (³H) was the most dominant radionuclide released in both gaseous and liquid effluents for all sites. Although significant amounts of ¹⁴C and noble gases such as ⁴¹Ar, ⁸⁵Kr, and ¹³³Xe were occasionally emitted. Among gaseous effluents, ³H, ¹⁴C, ⁴¹Ar, ¹³³Xe, ¹³¹ᵐXe, and ⁸⁵Kr showed high frequencies. For liquid effluents, ³H, ⁶⁰Co, ⁵⁸Co, and ⁹⁵Nb were the most frequently released. In terms of dose contribution, ¹⁴C, ³H, ¹³³Xe, ⁴¹Ar, and ⁶⁰Co were dominant among gaseous effluents, while ¹⁴C, ³H, ⁹⁵Nb, ⁵⁸Co, and ⁶⁰Co were significant among liquid effluents. The results of this study can serve as reference data for predicting future release trends and provide a basis for selecting key radionuclides for validation of the E-DOSE program.

Gaseous effluent, Liquid effluent, Nuclear power plant, Software validation & verification

During the normal operation of a nuclear power plant, radioactive effluents are released. Radioactive effluents can cause exposure to the public. To calculate dose, Korea Hydro & Nuclear Power (KHNP) is currently using the K-DOSE60, and developing new off-site dose calculation program E-DOSE, which has some improvements to the dose assessment methodology. Therefore, in this study, the dose assessment results of the two dose calculation codes, E-DOSE and K-DOSE60, were compared and analyzed. The dose of each pathway for gaseous effluents showed that E-DOSE and K-DOSE60 calculated the same results for cloudshine and inhalation, but different for groundshine and ingestion. The dose of each exposure pathway for liquid effluents showed E-DOSE and K-DOSE60 calculated the same results for boating and swimming, but different for beachshine and ingestion. The difference in dose by groundshine and beachshine is due to the consideration of daughter nuclides, and in ingestion due to the updated dose assessment model of 14C and 3H. Finally, the dose of gaseous effluents was 28.6% lower for E-DOSE than K-DOSE60, and that of liquid effluents was 10.5% higher for E-DOSE than K-DOSE60. The results of this study can be used for the development and software verification and validation of E-DOSE.

E-DOSE, K-DOSE60, Public dose, Offsite-dose calculation

Chelating agents, particularly aminopolycarboxylic acids (APCAs) such as ethylenediaminetetraacetic acid (EDTA), diethylenetriaminepentaacetic acid (DTPA), and nitrilotriacetic acid (NTA), are extensively employed across various industries, including agriculture, metallurgy, food processing, pharmaceuticals, and nuclear sectors. This is largely due to their exceptional ability to bind metal ions, effectively reducing their reactivity. However, this strong affinity also poses significant environmental challenges, including the potential for persistent contamination in aquatic systems. When used for chelating radioactive contaminants, the resulting complexes exhibit notable stability, potentially leading to prolonged environmental presence and radiological pollution risks. This study employs UV-Vis spectrophotometry to investigate the speciation of Ni-APCA complexes across different concentrations and pH levels, revealing insights into their structural dynamics and interactions. We found that while Ni-EDTA complexes are highly stable across a wide pH range, Ni-NTA and Ni-DTPA complexes exhibit significant structural variability. Understanding these dynamics is crucial for developing effective nuclear waste management strategies and enhancing analytical methods for APCA quantification. This research offers foundational knowledge to optimize APCA applications in decontamination processes.

Aminopolycarboxylic acids (APCAs), Nickel complexation, UV-Vis spectrophotometry, pH-dependent speciation

The safe stabilization of damaged spent nuclear fuel (SNF) is essential for its long‑term management and disposal. This study examines how burn‑up level, composition, and sintering temperature influence the microstructure and aqueous durability of stabilized ceramic pellets fabricated from surrogate SNF powders. Surrogates representing high burn‑up conditions (35 and 55 GWd/tU) were sintered at 1550 °C or 1700 °C, then ground and tested using the Product Consistency Test A at 90 °C for 7 days. Higher‑temperature sintering produced denser microstructures with smoother grains and lower porosity, while higher burn‑up increased surrogate content and slightly reduced densification. Leaching tests showed substantially reduced release of uranium and fission product surrogates from pellets sintered at 1700 °C, with each element’s behavior reflecting its microstructural association. ε‑phase particles enriched in Mo, Ru, and Pd were found to be highly resistant to dissolution. These results highlight the critical role of high‑temperature densification in enhancing chemical durability and support the use of stabilized pellets as robust waste forms for immobilizing damaged SNF, as represented by surrogate compositions in this study.

Stabilized pellets, Spent nuclear fuel, Burn‑up effects, Sintering temperature, Aqueous durability

Deep geological disposal of high-level radioactive waste requires ensuring long-term safety for humans and the environment over hundreds of thousands of years. Consequently, establishing a comprehensive hydrogeochemical monitoring framework is essential to quantitatively understand the disposal environment and track its long-term evolution. This study reviews the monitoring practices of leading disposal countries, focusing on three key aspects namely disposal stages, spatial scales, and hydrologic systems, and compares them with national monitoring systems in South Korea. Applying a consistent set of hydrogeochemical parameters across these aspects is essential for building reliable datasets that support safety assessments and radionuclide transport modeling. However, South Korea’s current monitoring systems operate separately for surface water and groundwater, with differing objectives and parameters, limiting their applicability for integrated analysis. To address this issue, a comprehensive monitoring plan based on existing networks must be developed. In addition, ensuring data reliability through a systematic quality assurance and quality control framework-covering monitoring design, sampling, laboratory analysis, and data management-is crucial for maintaining the scientific credibility of the monitoring results. This study provides principal insights and directions for the establishment of a hydrogeochemical monitoring system for future deep geological disposal in South Korea.

Deep geological disposal, Hydrogeochemical monitoring, Monitoring strategies, Disposal stages, Spatial scales, Hydrologic systems

Compacted bentonite buffer material plays a critical role in engineered barrier systems (EBSs) for high-level radioactive waste (HLW) disposal. Upon groundwater infiltration, the micro- and macropores within the bentonite structure undergo changes that induce swelling. Therefore, understanding the swelling behavior and hydraulic properties of compacted bentonite, considering both micro- and macropores contributions, is essential. In this study, the micro- and macroporosity of Bentonil-WRK bentonite, which has not been previously characterized, were determined using X-ray diffraction analysis. Additionally, the saturated hydraulic conductivity of compacted Bentonil-WRK bentonite was measured at various dry densities and compared with predictions from a modified Kozeny–Carman model that incorporates both micro- and macropore effects. The modified model accurately predicted saturated hydraulic conductivity, especially under low dry density conditions, highlighting the significance of considering macroporosity.

Bentonil-WRK bentonite, Basal spacing, Microporosity, Macroporosity, Saturated hydraulic conductivity, Kozeny–Carmen equation

In this research, a detailed analysis of the decay heat contributions of both actinides and non-actinides (fission fragments) from spent nuclear fuel (SNF) was made after 50 GWd·tHMM⁻¹ burnup of fresh uranium fuel with 4.5% enrichment lasted for 1,350 days. The calculations were made for a long storage period of 300 years divided into four sections 1, 10, 100, and 300 years so that we could study the decay heat and physical disposal ratios of radioactive waste in medium- and long-term storage periods. Fresh fuel burnup calculations were made using the code MCNP, while isotopic content and then decay heat were calculated using the built-in stiff equation solver in the MATLAB code. It is noted that only around 12 isotopes contribute more than 90% of the decay heat at all times. It is also noted that the contribution of actinides persists and is the dominant ether despite decreasing decay heat, while the effect of fission products decreases at a very rapid rate after about 40 years of storage.

SNF, Decay heat, Actinides, Fission fragments, Burnup, Stiff equations

The bentonite buffer material is a crucial component in an engineered barrier system used for the disposal of high-level radioactive waste. Because a large amount of heat from the disposal canister is released into the bentonite buffer material, the thermal conductivity of the bentonite buffer is a crucial parameter that determines the design temperature. At the Korea Atomic Energy Research Institute (KAERI), a new standard bentonite (Bentonil-WRK) has been used since 2022 because Gyeongju (KJ) bentonite is no longer produced. However, the currently available data are insufficient, making it essential to investigate both the basic and complex properties of Bentonil-WRK. Thus, this study evaluated its geotechnical and thermal properties and developed a thermal conductivity empirical model that considers its dry density, water content, and temperature variations from room temperature to 90°C. The coefficient of determination (R²) for the model was found to be 0.986. The thermal conductivity values of Bentonil-WRK were 1–10% lower than those of KJ bentonite and 10–40% higher than those of MX-80 bentonites, which were attributable to mineral-composition differences. The thermal conductivity of Bentonil-WRK ranged between 0.504 and 1.149 W·(m⁻¹·K⁻¹), while the specific heat capacity varied from 0.826 to 1.138 (kJ·(kg⁻¹·K⁻¹)).

Bentonil-WRK bentonite, Thermal conductivity, Specific heat capacity, Regression analysis

Hydraulic conductivity is a critical design parameter for buffers in high-level radioactive waste repositories. Most employed prediction models for hydraulic conductivity are limited to various types of bentonites, the main material of the buffer, and the associated temperature conditions. This study proposes the utilization of a novel integrated prediction model. The model is derived through theoretical and regression analyses and is applied to all types of compacted bentonites when the relationship between hydraulic conductivity and dry density for each compacted bentonite is known. The proposed model incorporates parameters such as permeability ratio, dynamic viscosity, and temperature coefficient to enable accurate prediction of hydraulic conductivity with temperature. Based on the results obtained, the values are in good agreement with the measured values for the selected bentonites, demonstrating the effectiveness of the proposed model. These results contribute to the analysis of the hydraulic behavior of the buffer with temperature during periods of high-level radioactive waste deposition.

Hydraulic conductivity, Compacted bentonite, Buffer, Prediction model, Engineered barrier system

During the dismantling of nuclear facilities, a large quantity of radioactive concrete is generated and chelating agents are required for the decontamination process. However, disposing of environmentally persistent chelated wastes without eliminating the chelating agents might increase the rate of radionuclide migration. This paper reports a rapid and straightforward ion chromatography method for the quantification of citric acid (CA), a commonly used chelating agent. The findings demonstrate acceptable recovery yields, linearities, and reproducibilities of the simulated samples, confirming the validity of the proposed method. The selectivity of the proposed method was confirmed by effectively separating CA from gluconic acid, a common constituent in concretes. The limits of detection and quantification of the method were 0.679 and 2.059 mg·L⁻¹, respectively, while the recovery yield, indicative of the consistency between theoretical and experimental concentrations, was 85%. The method was also employed for the quantification of CA in a real concrete sample. These results highlight the potential of this approach for CA detection in radioactive concrete waste, as well as in other types of nuclear wastes.

Chelating agent, Ion chromatography, Citric acid, Gluconic acid, Radioactive concrete waste, Quantitative analysis

In 2017, a decision was made to permanently shut down Kori Unit 1, and preparations began to be made for its decontamination and decommissioning. The dismantling of the biological shields concrete, reactor vessel (RV), and reactor vessel internals (RVI) is crucial to the nuclear decommissioning process. These components were radiologically activated by the neutron activation reaction occurring in the reactor during its operational period. Because of the radioactivity of the RV and RVI of Kori Unit 1, remotely controlled systems were developed for cutting within the cavity to reduce radiation exposure. Specialized equipment was developed for underwater cutting operations. This paper focuses on modeling related to RVI operations using the MAVRIC code and the dose calculation for a diver entering the cavity. The upper and lower parts of the RVI are classified as low-level radioactive waste, while the sides that came into contact with the fuel are classified as intermediate-level radioactive waste. Therefore, the modeling presented in this paper only considers the RVI sides because the upper and lower parts have a minimal impact on the radiation exposure. These research findings are anticipated to contribute to enhancing the efficiency and safety of nuclear reactor decommissioning operations.

MAVRIC, RVI, Diver, Kori 1, Dose, Radiation exposure

Currently, off-site dose calculations for nuclear power plants are conducted using a computer program (K-DOSE 60). The program is developed based on the regulatory guidelines of the Korea Institute of Nuclear Safety (KINS), which is a domestic nuclear regulatory agency. In this study, a domestic application of the International Atomic Energy Agency (IAEA) TRS (Technical Reports Series)-472 methodology for ³H and ¹⁴C in liquid effluents was studied. The dose-evaluation methods adopted and the program configuration for dose evaluation are described based on ³H and ¹⁴C in the liquid-effluent-evaluation module of the computer program. The accuracy of the program is verified by comparing the program-calculated results with hand calculation values. Furthermore, a comparative evaluation with LADTAP II, which is a liquid-effluent-evaluation methodology developed by the U.S. NRC (Nuclear Regulatory Commission), is performed. The result confirms that the program-calculated results for the IAEA TRS-472 methodology are consistent with the hand calculation values. Meanwhile, the result of comparative evaluation with LADTAP II indicates different results depending on the methodology used.

Liquid effluent, IAEA TRS-472, Off-site dose calculation (ODC), Tritium, ¹⁴C

In this study, we investigated the suppression of the corrosion of cast iron in a copper–cast iron double-layered canister under local corrosion of the copper layer. The cold spray coating technique was used to insert metals with lower galvanic activity than that of copper, such as silver, nickel, and titanium, between the copper and cast iron layers. Electrochemically accelerated corrosion tests were performed on the galvanic specimens in KURT groundwater at a voltage of 1.0 V for a week. The results revealed that copper corrosion was evident in all galvanic specimens of Cu–Ag, Cu–Ni, and Cu–Ti. By contrast, the copper was barely corroded in the Cu–Fe specimens. Therefore, it was concluded that if an inactive galvanic metal is applied to the areas where local corrosion is concerned, such as welding parts, the disposal canister can overcome local or non-uniform corrosion of the copper canister for long periods.

Copper, Galvanic corrosion, Disposal canister, Cold spray coating, Double layer coating

Fundamental aspects of creating passivation layers for corrosion resistance in nuclear engineering applications, specifically the ability to form complete layers versus porous ones, are being explored in this study. Utilizing a laser ablation technique, 1,064 nm fire at 10 Hz with 60 pulses per shot and 0.5 mm between impact points, aluminum samples are treated in an attempt to create a fully formed passivation layer that will be tested in a LiCl-KCl eutectic salt. By placing these samples into an electrochemical environment mimicking a pyroprocessing system, corrosion rates, resistances and material characteristics are tested for one week and then compared between treated and untreated samples. In initial testing, linear sweep voltammetry indicates corrosion current density for the untreated sample at −0.038 mA·cm⁻² and treated samples at −0.024 mA·cm⁻² and −0.016 mA·cm⁻², respectively. This correlates to a control sample corrosion rate of −0.205 mm·yr⁻¹ and treated rates of −0.130 mm·yr⁻¹ and −0.086 mm·yr⁻¹ for samples 1 and 2. In addition, electrochemical impedance spectroscopy circuits show application of a longer-lasting porous passivation layer on the treated metal, compared to the naturally forming layer. However, the current technique fails to create a uniform protection layer across the sample.

Molten salt, Corrosion, Electrochemistry, Passivation, Aluminum, Laser ablation

Highly radioactive waste is solidified to restrict leaching, retain its shape, and maintain its structural stability to prevent it from affecting humans and the environment as much as possible. This operation should be performed consistently regardless of whether the waste is homogeneous or heterogeneous. However, currently, there are no specific performance requirements for heterogeneous waste in Korea. This study reviewed domestic research results and the status of overseas applications, and proposed immobilization requirements for heterogeneous waste to be applied in Korea. IAEA safety standards, domestic laws, and waste acceptance criteria were reviewed. The status of heterogeneous waste immobilization in countries such as the United States, France, and Spain was reviewed. Most countries treat heterogeneous waste by encasing it in concrete, and impose immobilization requirements on this concrete. Based on these data, safety standards for the thickness, compressive strength, and diffusion limit of this concrete material were proposed as immobilization requirements for heterogeneous waste disposal in Korea. Quantitative values for the above requirements need to be derived through quantitative assessments based on the characteristics of domestic heterogeneous waste and disposal facilities.

Heterogeneous waste, Immobilization, Performance requirement, Waste acceptance criteria, Concrete lining

Safe radiation management is essential not only for operational nuclear power plants but also for nuclear plants to be decommissioned. When spent nuclear fuel is present on-site, meticulous radiation emergency plans are necessary to ensure safety. In Korea, numerous radiation emergency plans have been established for operational nuclear reactors. These plans delineate distinct response mitigation measures for white, blue, and red emergencies. However, clear regulations are yet to be devised for radiation emergency plans for reactors to be decommission. Therefore, this study investigated the decommissioning plan and status of Kori unit 1 to comprehensively analyze the current status of decommissioning safety in Korea. In this study, radiation emergency plans of decommissioning nuclear power plants abroad were reviewed to confirm radiation emergency action levels. Furthermore, radioactive waste treatment facilities, to be used for decommissioning reactors in Korea were evaluated. Moreover, the study assessed emergency plans (especially, emergency initiating conditions) for operational nuclear power plants in Korea for potential use in the decommissioning phase. This study proposed an emergency initiating condition that can be used for decommissioning reactors in Korea. Considering the anticipated introduction of plasma torch melting facility in Korea, this study examined the conditions of radiation emergency plans can be altered. This study identified effective measures and guidelines for managing radiological emergency initiating conditions, and effective decommissioning of nuclear power plants in Korea.

Decommissioning, Radiological emergency plan, Radiological emergency levels, Emergency action levels, Emergency initiating conditions

In this study, we employed a small-scale experiment to demonstrate the introduction of a thin copper heat dissipation plate into a bentonite buffer layer of an engineered barrier system. This experiment designed for spent nuclear fuel disposal can effectively reduce the maximum temperature of the bentonite buffer layer, and ultimately, make it possible to reduce the area of the disposal site. For the experiment, a small-scale engineered barrier system with a copper heat dissipation plate was designed and manufactured. the thickness of the cylindrical buffer was about 2 cm, which was about 1/20 of KAERI Repository System (KRS). At a power supply of 250 W, the maximum buffer temperature reduced to a mere 1.8°C when the thin copper plate was introduced. However, the maximum surface temperature reduced to a remarkable 9.1°C, when a U-collar copper plate was introduced, which had a good contact with the other barrier layers. Consequently, we conclude that the introduction of the thin copper plate into the engineered barrier system for spent nuclear fuel disposal can effectively reduce the maximum buffer temperature in high-level radioactive waste disposal repositories.

Disposal canister, Heat dissipation, Copper plate, Thermal conductivity, Bentonite buffer

The 300 concrete silo systems installed and operated at the site of Wolsong nuclear power plant (NPP) have been storing CANDU spent nuclear fuel (SNF) under dry conditions since 1992. The dry storage system must be operated safely until SNF is delivered to an interim storage facility or final repository located outside the NPP in accordance with the SNF management policy of the country. The silo dry storage system consists of a concrete structure, liner steel plate in the inner cavity, and fuel basket. Because the components of the silo system are exposed to high energy radiation owing to the high radioactivity of SNF inside, the effects of irradiation during long-term storage must be analyzed. To this end, material specimens of each component were manufactured and subjected to irradiation and strength tests, and mechanical characteristics before and after irradiation were examined. Notably, the mechanical characteristics of the main components of the silo system were affected by irradiation during the storage of spent fuel. The test results will be used to evaluate the long-term behavior of silo systems in the future.

Dry storage system, Concrete silo, Irradiation, Mechanical property, Long-term behavior

To non-destructively determine the burnup of a spent nuclear fuel assembly, it is essential to analyze the nuclear isotopes present in the assembly and detect the neutrons and gamma rays emitted from these isotopes. Specifically, gamma-ray measurement methods can utilize a single radiation measurement value of ¹³⁷Cs or measure based on the energy peak ratio of Cs isotopes such as ¹³⁴Cs/¹³⁷Cs and ¹⁵⁴Eu/¹³⁷Cs. In this study, we validated the extent to which the results of gamma-ray measurements using cadmium zinc telluride (CZT) sensors based on ¹³⁷Cs could be accurately simulated by implementing identical conditions on MCNP. To simulate measurement scenarios using a lead collimator, we propose equations that represent radiation behavior that reaches the detector by assuming “Direct hit” and “Penetration with attenuation” situations. The results obtained from MCNP confirmed an increase in measurement efficiency by 0.47 times when using the CZT detector, demonstrating the efficacy of the measurement system.

Spent nuclear fuel, Burnup, MCNP, CZT, Reaction rate

Concrete structures must maintain their shielding abilities and structural integrity over extended operational periods. Despite the widespread use of dry storage systems for spent nuclear fuel, research on the properties of deteriorated concrete and their impact on structural performance remains limited. To address this significant research gap, static and dynamic material testing was conducted on concrete specimens carefully extracted from the outer wall of the High-flux Advanced Neutron Application ReactOr (HANARO), constructed approximately 30 years ago. Despite its age, the results reveal that the concrete maintains its structural integrity impressively well, with static compression tests indicating an average compressive strength exceeding the original design standards. Further dynamic property testing using advanced high-speed material test equipment supported these findings, showing the consistency of dynamic increase factors with those reported in previous studies. These results highlight the importance of monitoring and assessing concrete structures in nuclear facilities for long-term safety and reliability.

Dry storage system, Concrete structures, Deterioration, Dynamic material properties

Undeclared nuclear activities are challenging given the lack of information from the sites involved in such activities. Wide-area environmental sampling (WAES) can be an effective method to detect undeclared nuclear activities. However, it is crucial to address the potential risks during the WAES, including sample tampering or extortions. Therefore, tracking and monitoring of various on-site data is imperative to accurately interpret the status of samples and workers throughout the WAES process. ‘Environmental and Geographical Data Transfer (EGDT)’ was developed for the real-time monitoring of integrated on-site data. EGDT module is equipped with various sensors and can be attached to a worker’s uniform or a sample storage box. This study demonstrated the technical effectiveness of EGDT by exploring three experimental methodologies for feasibility assessment. Compared to the Normal Operation case, the inference of the Sample Extortion case was predominantly based on changes in lux and dose rate. The inference of the Out-of-Work-Area case primarily relied on changes in dose rate and acceleration. Finally, the preliminary evaluation of the performance of the developed prototype was conducted, and a foundation was established for enhancing the application in the WAES process.

Wide-area environmental sampling, Real-time data transfer, Nuclear activity detection, Undeclared area, Continuity of knowledge

This study aimed to develop an efficient recycling process for wastewater generated from soil-washing used to remediate uranium (U(VI))-contaminated soil. Under acidic conditions, U(VI) ions leached from the soil were precipitated and separated through neutralization using hydrazine (N₂H₄). N₂H₄, employed as a pH adjuster, was decomposed into nitrogen gas (N₂), water (H₂O), and hydrogen ions (H⁺) by hydrogen peroxide (H₂O₂). The residual N₂H₄ was precipitated when the pH was adjusted using sulfuric acid (H2SO₄) to recycle the wastewater in the soil-washing process. This purified wastewater was reused in the soil-washing process for a total of ten cycles. The results confirmed that the soil-washing performance for U(VI)-contaminated soil was maintained when using recycled wastewater. All in all, this study proposes an efficient recycling process for wastewater generated during the remediation of U(VI)-contaminated soil.

Soil-washing, Uranium precipitation, Hydrazine decomposition, Wastewater treatment, Water recycling

Vitrification, one of the most promising solidification processes for various materials, has been applied to radioactive waste to improve its disposal stability and reduce its volume. Because the thermal decomposition of dry active waste (DAW) significantly reduces its volume, the volume reduction factor of DAW vitrification is high. The KHNP developed the optimal glass composition for the vitrification of DAW. Since vitrification offers a high-volume reduction ratio, it is expected that disposal costs could be greatly reduced by the use of such technology. The DG-2 glass composition was developed to vitrify DAW. During the maintenance of nuclear power plants, metals containing paper, clothes, and wood are generated. ZrO₂ and HfO₂ are generally considered to be network-formers in borosilicate-based glasses. In this study, a feasibility study of vitrification for DAW that contains metal particulates is conducted to understand the applicability of this process under various conditions. The physicochemical properties are characterized to assess the applicability of candidate glass compositions.

Vitrification, DAW, Solidification, Compressive strength, PCT

Disposal cover as an engineered barrier of a near-surface disposal facility for low and very low-level radioactive waste is composed of a multi-layer to isolate radioactive waste from environmental influences for the long term. To acquire a realistic forecast for the post-closure period of the disposal facility, it is essential to carry out long-term experimental research in a similar condition to the actual disposal environment. Hence, a performance test facility of the disposal cover was constructed in Gyeongju low and intermediate level radioactive waste disposal center in 2022. The constructed performance test facility has differences from the material properties presented in the design. These differences are factors that affect the prevent rainfall infiltration, which is one of the important roles of the disposal cover. Therefore, in this study, a numerical simulation of rainfall infiltration into the performance test facility was performed for the designed case and the actual constructed case. To simulate the behavior of water infiltration, the FEFLOW software based on the finite element method is used. Through the analysis of numerical simulation results, it is confirmed that the hydraulic conductivity of the material constituting the multi-layer of the disposal cover greatly influences the amount of water infiltration.

Low and intermediate-level radioactive waste, Disposal cover, Performance test facility, Rainfall infiltration, Near-surface disposal facility

Because most spent nuclear fuel storage casks have been designed for low burnup fuel, a safety-significant high burnup dry storage cask must be developed for nuclear facilities in Korea to store the increasing high burnup and damaged fuels. More than 20% of fuels generated by PWRs comprise high burnup fuels. This study conducted a structural safety evaluation of the preliminary designs for a high burnup storage cask with 21 spent nuclear fuels and evaluated feasible loading conditions under normal, off-normal, and accident conditions. Two types of metal and concrete storage casks were used in the evaluation. Structural integrity was assessed by comparing load combinations and stress intensity limits under each condition. Evaluation results showed that the storage cask had secured structural integrity as it satisfied the stress intensity limit under normal, off-normal, and accident conditions. These results can be used as baseline data for the detailed design of high burnup storage casks.

Storage cask, High burnup spent nuclear fuel, Damaged fuel, Tip-over

The growing significance of sustainable energy technologies underscores the need for safe and efficient management of spent nuclear fuels (SNFs), particularly via deep geological disposal (DGD). DGD involves the long-term isolation of SNFs from the biosphere to ensure public safety and environmental protection, necessitating materials with high corrosion resistance for DGD canisters. This study investigated the feasibility of a Cu–Ni film, fabricated via additive manufacturing (AM), as a corrosion-resistant layer for DGD canister applications. A wire-fed AM technique was used to deposit a millimeter-scale Cu–Ni film onto a carbon steel (CS) substrate. Electrochemical analyses were conducted using aerated groundwater from the KAERI underground research tunnel (KURT) as an electrolyte with an NaCl additive to characterize the oxic corrosion behavior of the Cu–Ni film. The results demonstrated that the AM-fabricated Cu–Ni film exhibited enhanced corrosion resistance (manifested as lower corrosion current density and formation of a dense passive layer) in an NaCl-supplemented groundwater solution. Extensive investigations are necessary to elucidate microstructural performance, mechanical properties, and corrosion resistance in the presence of various corroding agents to simplify the implementation of this technology for DGD canisters.

Copper–nickel film, Additive manufacturing, Corrosion, Deep geological disposal canister

Democratic People’s Republic of Korea (DPRK) has produced weapon-grade plutonium in a graphite-moderated experimental reactor at the Yongbyon nuclear facilities. The amount of plutonium produced can be estimated using the Graphite Isotope Ratio Method (GIRM), even without considering specific operational histories. However, the result depends to some degree on the operational cycle length. Moreover, an optimal cycle length can maximize the number of nuclear weapons made from the plutonium produced. For conservatism, it should be assumed that the target reactor was operated with an optimal cycle length. This study investigated the optimal cycle length using which the Calder Hall MAGNOX reactor can achieve the maximum annual production of nuclear weapons. The results show that lower enrichment fuel produced a greater number of critical plutonium spheres with a shorter optimal cycle length. Specifically, depleted uranium (0.69wt%) produced 5.561 critical plutonium spheres annually with optimal cycle lengths of 251 effective full power days. This research is crucial for understanding DPRK’s potential for nuclear weapon production and highlights the importance of reactor operational strategy in maximizing the production of weapons-grade plutonium in MAGNOX reactors.

MAGNOX, Weapon-grade plutonium production, DPRK nuclear capabilities, Cycle length optimization, GIRM

After the permanent shut down of Kori Unit 1, various decommissioning activities will be implemented, including decontamination, segmentation, waste management, and site restoration. During the decommissioning period, waste management is among the most important activities to ensure that the process proceeds smoothly and within the expected timeframe. Furthermore, the radioactive waste generated during the operation should be sent to a disposal facility to complete the decommissioning project. Square and cylindrical concrete re-package drums were generated during the 1980s and 1990s. The square, containing boron concentrates, and cylindrical, containing spent resin, concrete re-package drums have been stored in a radioactive waste storage building. Homogeneous radioactive waste, including boron concentrates, spent resin, and sludge, should be solidified or packaged in high-integrity containers (HICs). This study investigates the sequential segmentation process for the separation of contaminated and non-contaminated regions, the re-packaging process of segmented or crushed cement-solidified boron concentrate, and re-packaging in HICs. The conceptual design evaluates the re-packaging plan for the segmented and crushed cement-solidified waste using HICs, which is acceptable in a disposal facility, and the quantity of generated HICs from the treatment process.

Boron concentrates, Spent resin, Large concrete drum, High-integrity container, Low level waste

The initial radionuclide migration quantity depends on the total amount of solubilized species. Geochemical modeling based on a thermodynamic database (TDB) has been employed to assess the solubility of radionuclides. It is necessary to evaluate whether the TDB describes the domestic repository conditions appropriately. An effective way to validate the TDB-based modeling results is through direct comparisons with experimentally measured values under the conditions of interest. Here, the solubilities of trivalent Sm, Eu, and Am were measured in synthetic KURT-DB3 groundwater (SynDB3) and compared with modeling results based on ThermoChimie TDB. Ln₂(CO₃)₃·xH₂O(cr) (Ln = Sm, Eu) solids were introduced into the Syn-DB3 and dissolved Sm and Eu concentrations were monitored over 223 days. X-ray diffraction analysis confirmed that the crystallinity of the solid compounds was maintained throughout the experiments. The dissolved Sm and Eu concentrations at equilibrium were close to the predicted solubilities of Sm₂(CO₃)₃(s) and Eu₂(CO₃)₃(s) based on the ThermoChimie TDB. The Am solubility measured under oversaturated conditions was comparable to the measured Eu concentrations, although they were measured under different experimental settings. More experimental data are needed for Am-carbonate solid systems with careful characterization of the solid phases to better evaluate Am solubility in domestic groundwater conditions.

Trivalent americium, Trivalent lanthanide, Solubility, KURT groundwater, Radionuclide migration, Thermodynamic database

Pyrochemical processing and molten-salt reactors have recently garnered significant attention as they are promising options for future nuclear technologies, such as those for recycling spent nuclear fuels and the next generation of nuclear reactors. Both of these technologies require the use of high-temperature molten salt. To implement these technologies, one must understand the electrochemical behavior of fission products in molten salts, lanthanides, and actinides. In this study, a rotating-disk-electrode (RDE) measurement system for high-temperature molten salts is constructed and tested by investigating the electrochemical reactions of Sm³⁺ in LiCl–KCl melts. The results show that the reduction of Sm³⁺ presents the Levich behavior in LiCl–KCl melts. Using the RDE system, not only is the diffusion-layer thickness of Sm³⁺ measured in high-temperature molten salts but also various electrochemical parameters for Sm³⁺ in LiCl–KCl melts, including the diffusion coefficient, Tafel slope, and exchange current density, are determined.

Pyrochemical processing, LiCl−KCl, Rotating disk electrode, Samarium, Tafel slope, Exchange current density

A procedure for minimizing the environmental burden and maximizing the efficiency of storage sites used for the final disposal of spent fuel has been proposed. In this procedure, fission products (highly mobile and producing heat) are collected, and uranium and TRU-RE (transuranium-rare earth) oxide are independently stored. The possibility and applicability of radiation measurement for monitoring the nuclear materials effectively throughout the process has been simulated and evaluated. For the simulation, the properties of the chemical processes were analyzed, the major radiation emitters were determined, and the production of nuclear materials by chemical reactions were evaluated. In each process, the content of nuclear material was changed by up to 20% to represent abnormal conditions. The results showed that the plutonium peak was matched with the change in the TRU content and the measured signal was changed linearly with respect to the content change of the plutonium. From the neutron measurement, a linear response of the TRU content variation was obtained. In addition, a logic diagram was developed for the nuclear monitoring. The integration of radiation detections is recommended for monitoring the process effectively and efficiently.

Nuclear material, Monitoring, Fuel process, Measurement, Radiation, Spent fuel

Technetium-99 is identified as an element of interest for the safety assessment of a deep geological repository for used nuclear fuel. The sorption behavior of Tc(IV) onto MX-80 and granite in Ca-Na-Cl solutions of varying ionic strength (0.05–1 mol·kgw⁻¹ (m)) and across a pHm range of 4–9 was studied in this paper. Sorption of Tc(IV) was found to be independent of ionic strength in the range of 0.05 to 1 m for both MX-80 and granite. Sorption of Tc(IV) on MX-80 increased with pH_m from 4 to 7 and then decreased with pH_m from 8 to 9. Sorption of Tc(IV) on granite gradually increased with pH_m from 4 to 8 and then became almost constant or slightly decreased with pH_m from 8 to 9. A 2 site protolysis non-electrostatic surface complexation and cation exchange sorption model successfully simulated sorption of Tc(IV) on MX-80 and granite. Optimized values of surface complexation constants (log K⁰) are proposed.

Technetium(IV), Sorption, Granite, MX-80, Saline, 2SPNE SC/CE model

Technetium has been identified as an element of interest for the safety assessment of a deep geological repository for used nuclear fuel. In this study, the sorption of Tc(IV) onto MX-80 bentonite, illite, and shale in ionic strength (I) 0.1–6 mol·kgw⁻¹ (m) Na-Ca-Cl solutions at pH_m = 4–9 and limestone at pH_m = 5–9 was studied. Tc(IV) sorption on MX-80 increased with pH_m from 4 to 6, reached the maximum at pH_m = 6–7, and then gradually decreased with pH_m from 7 to 9. Tc(IV) sorption on illite gradually increased with pH_m from 4 to 7, and then decreased as pHm increased. The sorption properties of Tc(IV) on shale were quite similar to those on illite. Tc(IV) sorption on limestone slightly increased with pH_m from 5 to 6 and then seemed to be constant at pH_m = 6–9. Tc(IV) sorption on all four solids was independent of ionic strength (0.1–6 m). The 2 site protolysis non-electrostatic surface complexation and cation exchange model successfully simulated the sorption of Tc(IV) onto MX-80 and illite and the optimized values of surface complexation constants were estimated.

Technetium(IV), Sorption, MX-80 bentonite, Illite, Shale, Limestone

Hydride analysis is required to assess the mechanical integrity of spent nuclear fuel cladding. Image segmentation, which is a hydride analysis method, is a technique that can analyze the orientation and distribution of hydrides in cladding images of spent nuclear fuels. However, the segmentation results varied according to the image preprocessing. Inaccurate segmentation results can make hydride difficult to analyze. This study aims to analyze the segmentation performance of the Otsu algorithm according to the morphological operations of cladding images. Morphological operations were applied to four different cladding images, and segmentation performance was quantitatively compared using a histogram, betweenclass variance, and radial hydride fraction. As a result, this study found that morphological operations can induce errors in cladding images and that appropriate combinations of morphological operations can maximize segmentation performance. This study emphasizes the importance of image preprocessing methods, suggesting that they can enhance the accuracy of hydride analysis. These findings are expected to contribute to the advancements in integrity assessment of spent nuclear fuel cladding.

Hydride, Image segmentation, Morphological operations, Otsu algorithm, Spent nuclear fuel

Se sorption onto Ca-type montmorillonite purified from Bentonil-WRK—a new research bentonite introduced by Korea Atomic Energy Research Institute—was examined under ambient conditions (pH 4−9, pe 7−9, I = 0.01 M CaCl₂, and T = 25°C). Se(IV) was identified as the oxidation state responsible for weak sorption (Kd < 22 L∙kg⁻¹) by forming surface complexes with edge functional groups of the montmorillonite. Thermodynamic modeling, considering reaction mechanisms of outer-sphere complexation (≡AlOH₂⁺ + HSeO₃⁻ ⇌ ≡AlOH₃SeO₃, log K = 0.50 ± 0.21), inner-sphere complexation (2≡AlOH + H₂SeO₃(aq) ⇌ (≡Al)₂SeO₃ + 2H₂O(l), log K = 7.89 ± 0.51), and Ca²⁺-involved ternary complexation (≡AlOH + Ca²⁺ + SeO₃ ²⁻ ⇌ ≡AlOHCaSeO₃, log K = 7.69 ± 0.28) between selenite and aluminol sites of montmorillonite, acceptably reproduced the batch sorption data. Outer- and inner-sphere complexes are predominant Se(IV) forms sorbed in acidic (pH ≈ 4) and near-acidic (pH ≈ 6) regions, respectively, whereas ternary complexation accounts for Se(IV) sorption at neutral pHs under the ambient conditions. The experimental and modeling data generally extend a material-specific sorption database of Bentonil-WRK, which is essential for assessing its radionuclide retention performance as a buffer candidate of deep geological disposal system for high-level radioactive waste.

Bentonite, Montmorillonite, Bentonil-WRK, Selenium, Thermodynamic sorption modeling, Surface complexation

Interim dry cask storage systems comprising AISI 304 or 316 stainless steel canisters have become critical for the storage of spent nuclear fuel from light water reactors in the Republic of Korea. However, the combination of microstructural sensitization, residual tensile stress, and corrosive environments can induce chloride-induced stress corrosion cracking (CISCC) for stainless steel canisters. Suppressing one or more of these three variables can effectively mitigate CISCC initiation or propagation. Surface-modification technologies, such as surface peening and burnishing, focus on relieving residual tensile stress by introducing compressive stress to near-surface regions of materials. Overlay coating methods such as cold spray can serve as a barrier between the environment and the canister, while also inducing compressive stress similar to surface peening. This approach can both mitigate CISCC initiation and facilitate CISCC repair. Surface-painting methods can also be used to isolate materials from external corrosive environments. However, environmental variables, such as relative humidity, composition of surface deposits, and pH can affect the CISCC behavior. Therefore, in addition to research on surface modification and coating technologies, site-specific environmental investigations of various nuclear power plants are required

Dry cask storage system, Peening, Burnishing, Cold-spray coating

The spent nuclear fuel, combusted and released in the nuclear power plant, is stored in the spent fuel pool (SFP) located in the fuel buildings interconnected with the reactors. In Korea, spent fuel has been stored exclusively in SFPs, prompting initiatives to expand storage capacity by either installing additional SFPs or replacing them with high-density spent fuel storage racks. The installation of these fuel racks necessitates obtaining a regulatory license contingent upon ensuring safe fuel handling and storage systems. Regulatory agencies mandate the formulation of various postulated accident scenarios and assessments covering criticality, shielding, thermal behavior, and structural integrity to ensure safe fuel handling and storage systems. This study describes an evaluation method for assessing the structural damage to storage racks resulting from fuel dropping as a part of the functional safety evaluation of these racks. A scenario was envisaged wherein fuel was dropped onto the base plates of the upper and lower sections of the storage racks, and the impact load was analyzed using the ABAQUS/Explicit program. The evaluation results revealed localized plastic deformation but affirmed the structural integrity and safety of the storage racks.

Spent fuel rack, Shallow drop, Deep drop, Spent fuel drop

In Korea, two types of spent nuclear fuels (SNFs) are generated, pressurized light water reactor type (PWR) and pressurized heavy water reactor type (PHWR; CANDU), that differ greatly in size, decay heat, and radioactive characteristics. Technology development for the disposal of SNFs has mainly focused on PWR SNFs that are large in size and have extremely high decay heat and radioactivity. However, CANDU SNFs should be considered differently from PWR SNFs in deep geological disposal systems because their characteristics significantly differ from those of PWR SNFs in terms of their dimensions, number of SNF bundles, and handling systems in nuclear power plant sites. In this paper, after reviewing the status of the CANDU SNF disposal concept by Canada and Korea, concepts related to the direct geological disposal of CANDU SNFs were described, and two concepts were proposed based on the results of the development. The engineered barrier systems developed using these two concepts were comparatively analyzed in terms of disposal safety, disposal efficiency, and technical maturity. Based on the results of the comparative analyses, a vertical-type emplacement disposal concept was determined as a reference concept for the deep geological disposal of CANDU SNFs.

Spent nuclear fuels, CANDU bundles, Geological disposal concept, Engineered barrier system, Vertical emplacement, Horizontal emplacement, Reference disposal concept

Kori Unit 1, the first commercial nuclear power plant (NPP) in Korea, was permanently shut down in 2017 and was scheduled for decommissioning. Various programs must be planned early in the decommissioning process to safely decommission NPPs. Radiological characterization is a key program in decommissioning and should be a high priority. Radiological characterization involves determining the decommissioning technology to be applied to a nuclear facility by identifying the radiation sources and radioactive contaminants present within the facility and assessing the extent and nature of the radioactive contaminants to be removed from the facility. This study introduces the regulatory requirements, procedures, and implementation methods for radiological characterization and proposes a methodology to link the results of radiological characterizations for each stage. To link radiological characteristics, this study proposes to conduct radiological characterization in the decommissioning phase to verify the results of radiological characterization in the transitional phase of decommissioning NPPs. This enables significantly reducing the scope and content of radiological characterization that must be performed in the decommissioning phase and maintaining the connection with the previous phase.

Nuclear power plants, Decommissioning, Radiological characterization, Final decommissioning plans

The concrete silo dry storage system, which has been in operation at the Wolsong NPP site since 1992, consists of a concrete structure, a steel liner plate in the inner space, and a fuel basket. The silo system’s concrete structure must maintain structural integrity as well as adequate radiation shielding performance against the high radioactivity of spent nuclear fuel stored inside the storage system. The concrete structure is directly exposed to the external climatic environment in the storage facility and can be expected to deteriorate over time owing to the heat of spent nuclear fuel, as well as particularly cracks in the concrete structure. These cracks may reduce the radiation shielding performance of the concrete structure, potentially exceeding the silo system’s allowable radiation dose rate limits. For specimens with the same composition and physical properties as silo’s concrete structures, cracks were forcibly generated and then irradiated to measure the change in radiation dose rate to examine the effect of cracks in concrete structures on radiation shielding performance, and in the current state, the silo system maintains radiation shielding performance.

Dry storage system, Concrete silo, Radiation shielding, Crack, Surface dose rate

Increasing resource use is the primary motivator for the development of technology industries, which is leading to severe consequences, such as the release and disposal of radioactive waste containing radionuclides in the environment. Cesium (¹³⁷Cs) is one of the most hazardous radionuclides in the environment. In particular, the steel manufacturing process produces hazardous waste in the form of electric arc furnace dust contaminated with ¹³⁷Cs. In this study, the tolerances of five legume species to different activity concentrations of ¹³⁷Cs in both seed germination and initial seedling growth were compared. To determine ¹³⁷Cs tolerance, several parameters related to the growth and development of legumes were measured. Among the five legumes studied, Crotalaria juncea L. was the most ¹³⁷Cs tolerant at 50,000 Bq·L⁻¹. Sesbania javanica Miq. and Vigna mungo L. Hepper were moderately tolerant to 30,000 Bq·L^{−1 137}Cs. After 10 days, the stress tolerance indices in all legume species decreased by more than 50% at activity concentrations greater than or equal to 20,000 Bq·L^{−1 137}Cs. This approach allows the selection of desirable traits, making more-effective application possible in the phytoremediation of ¹³⁷Cs through stress tolerance. In conclusion, legumes are promising candidates for the phytoremediation of environmental pollutants.

Electric arc furnace dust, Legumes, Cesium (¹³⁷Cs), Seed germination

Industries that use or produce radionuclides have unintentionally released these substances into surrounding soils and sediments. To address this problem, Beautiful Environmental Construction (BEC) Co. developed the BEC’s Radioactive Soil Decontamination (BERAD) system to remove contaminants and reduce the volume of radionuclide-contaminated soils. Owing to the limited availability of radioactive isotopes such as ⁶⁰Co, ⁹⁰Sr, ¹³⁷Cs, and uranium-contaminated soil, naturally occurring elements in soil were used in this demonstration. The soil was divided into six size fractions via manual wet sieving and the BERAD system. Then, the concentrations of uranium, cobalt, strontium, cesium, and iron in each fraction were measured. The results of both separations showed that a considerable amount of each element is retained in the finer size fractions (<0.2 mm). After BERAD separation, the corresponding values yielded 53% uranium, 45% strontium, 66% cobalt, and 53% cesium in the fine size (<0.2 mm) fractions of 35% by weight. The study found that the concentrations of these elements increased as the particle sizes decreased. Iron and micaceous minerals played a significant role in retaining the elements. The pilot scale BERAD system yielded results that were similar to those obtained via laboratory wet-sieving and was successfully demonstrated as a soil washing technology.

Rad-contaminated soil, Remediation, Soil-washing, Uranium, Cesium, Cobalt, Strontium

The irradiated large components in a nuclear power plant must perform dismantling after performing a lot of training through mockup, but this mockup training is different the realistic environment such as a high radiation environment. Therefore, to develop a training system deployed in eXtended Reality (XR) that mimics a realistic environment, a suitable training and evaluation processes need to be developed. The target components for the training system select reactor, reactor internal and steam generator, and the training work of dismantling processes select dismantling, packaging and dismantling support work by reviewing the dismantling process of each component for XR application. By analyzing the detailed process of the selected dismantling process, Virtual Reality (VR) and Augmented Reality (AR) application processes will be developed. The basic contents of training system present purpose, goal, period, and evaluation for training and the training contents develop the basic, improvement and intensified course by level. The basic model for evaluation develop 3 stages (reaction, training, behavior) and evaluation process present detail training contents, target and evaluation method, evaluation criteria by level. Additionally, the developed training course and evaluation models will be useful for worker training in dismantling process of nuclear power plant.

Radioactive component, Extended reality, Reactor, Reactor internal, Dismantling process, Training system

A scintillator using organic materials can be easily manufactured in various shapes and sizes to suit the user’s purpose. A quantum dot (QD)-based scintillator has a number of advantages over commercial scintillators, including emission wavelength control, high-purity emission of a specific wavelength, high photoluminescence efficiency, and good photostability. The organic scintillators doping with various agents into the polymer media to increase scintillation efficiency and to control the emissioning wavelength through energy transfer process. In this study, scintillator enhancement was observed with different QDs material and detection response to gamma and neutron was investigated in energy spectrum. Multishell-structure QDs (CdS/CdZnS/ZnS) were fabricated and utilized to offset the shortcomings of single-shell-structure QDs, and the optical properties and the gamma and neutron detection performance capabilities were evaluated. The results of the evaluation of the detection response of the QD-based scintillator confirmed that the neutron/gamma classification performance was similar to that of a commercial scintillator. Furthermore, the gamma detection efficiency was improved by 34–38% (in the case of ¹³⁷Cs) compared to a commercial scintillator. This study is especially notable in that the organic scintillator incorporated with the newly fabricated QDs can be utilized for gamma and neutron detection for the operation and decommissioning various nuclear facilities.

Quantum dot (QD), Solid/Liquid scintillator, Detection efficiency, Gamma detection, Neutron/Gamma identification (PSD)

The recent release of contaminated water from the Fukushima Daiichi Nuclear Power Plant highlights the need for accurate tritium measurement, particularly near the minimum detectable activity (MDA) of 5 Bq·L⁻¹ set by South Korea’s Nuclear Safety and Security Commission. This study aims to improve low-level tritium measurement accuracy by optimizing the region of interest (ROI) for quench curve determination. These adjustments are crucial for separating tritium signals from background noise. Quench standards were prepared and measured using a liquid scintillation counter (LSC). Three ROIs were analyzed to assess the impact of channel selection on measurement precision: A 20-148 channel range optimized via figure of merit (FOM) analysis, a 20-250 channel range covering tritium’s full beta spectrum, and a broad 1-1024 channel range. Quench curves were obtained by fitting the counting efficiency of each ROI to the quench standards. Tritium samples with six different activity levels were prepared, and their radioactivity was calculated using the quench curves. Selecting appropriate ROIs for quench curve determination is critical for measuring low-concentration tritium accurately. This approach reduces uncertainty and emphasizes reliable methods to improve the precision and consistency of tritium measurements.

Tritium (³H), Liquid scintillation counter (LSC), Quench curve, Region of interest (ROI)

This study evaluates the long-term performance of a multi-layer cover system (MLCS) for near-surface disposal facilities using numerical modeling to estimate infiltration rates under various rainfall scenarios. An effective cover system is essential to prevent radionuclide migration and protect groundwater inflow within disposal facility. The analysis incorporated different bedrock characteristics (homogeneous and discrete fracture networks) and rainfall patterns throughout a 300-year post-closure period, assuming constant initial hydraulic properties. A comprehensive modeling approach incorporating both saturated and unsaturated flow dynamics was employed to assess system performance. Results showed that the cover system effectively limited infiltration rates to 15.94%−21.25% of the design criterion (32 mm∙year⁻¹) across all scenarios. Although infiltration patterns showed minimal sensitivity to bedrock heterogeneity, preferential flow along fractures was observed in the unsaturated zone, necessitating further investigation. These findings emphasize the importance of considering fracture-dominated flow in cover system design and highlight the need for detailed analysis of chemical degradation effects, experimental validation, and uncertainty quantification. The study provides valuable insights for optimizing disposal facility designs and improving long-term performance assessment methodologies.

Multi-layer cover system, Near-surface, Infiltration, Unsaturated zone, Fracture flow, Long-term performance, Numerical modeling

The increase in radioactive waste increased the demand for transportation to the disposal facility. Prior to transporting radioactive waste, confirming that the potential exposure is insignificant is crucial. Overland transportation risk assessment models were developed tailored to domestic characteristics. Dose assessment using this model requires selecting appropriate factors. However, users may struggle to derive appropriate values, leading to inaccuracies. Additionally, if assessment results show outliers, prioritizing factors for review can be challenging. Therefore, sensitivity analysis is necessary to prioritize factors for accurate assessment. In this study, sensitivity analysis was conducted on the on-link public risk assessment model factors for radioactive waste overland transportation. Initially, assessment models were analyzed by each detailed exposure scenario. Subsequently, uncertainty propagation-based sensitivity analysis methodology was applied. The default values for the assessment model factors were set, and sensitivity analysis was conducted based on road type for maximum individual and collective dose assessment models. For the maximum individual dose model, the distance to the samedirection vehicle was the most sensitive, whereas for the collective dose model, vehicle velocity was the most sensitive. The results of this study can be used as the basic data on radioactive waste transportation risk assessment in Korea in the future.

Radioactive waste, Overland transportation, Sensitivity analysis, Assessment model factor, On-link public

We analyzed the mineral composition of compacted calcium bentonite (GJ-I) and uncompressed sodium bentonite (MX80), both of which were exposed for two years in the YS03 borehole. The YS03 borehole is characterized by a high concentration of anaerobic microorganisms, including sulfate-reducing bacteria, elevated levels of hydrogen sulfide, and high pH conditions. The compacted Ca bentonite showed minimal alteration, with a small amount of new magnetite formation. However, an X-ray diffraction (XRD) analysis revealed that the uncompressed Na bentonite underwent a complete transformation from montmorillonite to muscovite, goethite, and magnetite. Therefore, it is suspected that the compactness of the bentonite significantly impacts the rate of alteration. Furthermore, an X-ray fluorescence (XRF) analysis demonstrated a marked increase in iron oxide in the Na bentonite, whereas key elements of montmorillonite such as alumina (Al₂O₃), silica (SiO₂), and magnesium oxide (MgO) showed substantial decreases. The presumed cause of the alteration in the uncompressed MX80 bentonite is the presence of Fe cations coupled with a high pH environment. We believe that Fe cations, which were likely released from the corrosion of cast iron, played a significant role in altering the montmorillonite lattice.

Bentonite, Montmorillonite, Alteration, Hydrogen sulfide, Muscovite, Goethite

A multi-barrier can be applied for the deep geological disposal of high-level radioactive waste. The multi-barrier comprises an engineered barrier and the natural barrier of the host rock. In the engineered barrier, the bentonite buffer is the key component for the disposal container, and the bentonite buffer thickness is given important consideration when designing the engineered barrier. This study reviewed the safety functions of bentonite buffers. Subsequently, the requirements and factors necessary to determine the thickness of the bentonite buffer, including criteria for radiological safety and the thermal stability of the disposal system, were identified. Additionally, the bentonite buffer thicknesses required for the top, bottom, and side of the disposal container were calculated. A double-layered emplacement method is also proposed for the bentonite buffer to improve disposal efficiency in terms of thermal management. Based on radiological safety and thermal stability analyses, an optimal thickness of 0.36 m was found to be appropriate for the bentonite buffer surrounding the disposal container. The thickness of the bentonite buffer above the disposal container can be determined based on the excavation damaged zone depth. The study findings can be used as a reference when designing deep geological disposal systems.

Deep geological disposal, Spent nuclear fuels, Buffer thickness, Engineered barrier, Radiological safety, Thermal stability

Deep geological repositories (DGR) count amongst the world largest environmental protection projects. They are the internationally advocated reference solution for the long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF). Many countries have engaged in programs to develop their own DGR. In Europe, four countries have passed the important milestone of choosing or announcing the appropriate site for the location of their HLW disposal facilities. Finland has almost finished the commissioning of its DGR and should start industrial disposal operations in 2025. Sweden has authorized the construction of its DGR and is finalizing the licensing document to start construction. France is reviewing the construction license application of its DGR. Switzerland has proposed the location of its DGR and is carrying out the production of the documentation for the license application. These four countries took decades to choose the location for their DGR. The length of this process is explained by 1) the amount of technical investigations and studies that were carried out to first identify, select and then fully characterize the suitable site and 2) the progressive decision-making process defined by their respective legal frameworks, including the participation and engagement of communities and stakeholders.

High-level radioactive waste, Spent nuclear fuel, Deep geological disposal, Underground research laboratories, Long-term environmental protection, Repository siting methodology

The transportation of spent nuclear fuel between management stages is expected, and the transportation workers may be exposed to radiation. When transporting spent nuclear fuel, the ALARA principle must be observed for the workers. The obj!ective of this study is to assess a radiation dose for workers transporting spent nuclear fuel using metal overpacks. For this obj!ective, the cask to be handled was selected and the radiation source term was set. Then, the radiation exposure scenario for the transportation workers was defined. Finally, the dose rates for each location of operation were assessed using Monte Carlo simulations, and collective doses were derived for each operation considering the radiation exposure scenario. Each worker performed 11 operations to transport spent nuclear fuel to other facilities and was exposed to a total of 1.138 man-mSv. The operation of removing the bottom shield ring resulted in the highest radiation exposure at 0.503 man-mSv. In contrast, the operation of installing the impact limiter resulted in the lowest radiation exposure at 0.0009 man-mSv. The results of this study can be used to strengthen radiation protection measures for workers transporting spent nuclear fuel in dry storage facilities using metal overpacks.

Transporting spent nuclear fuel procedure, Metal overpack, Occupational radiation dose, Spent nuclear fuel

The major concern in the deep geological disposal of spent nuclear fuels include sulfide-induced corrosion and stress corrosion cracking of copper canisters. Sulfur diffusion into copper canisters may induce copper embrittlement by causing Cu₂S particle formation along grain boundaries; these sulfide particles can act as crack initiation sites and eventually cause embrittlement. To prevent the formation of Cu₂S along grain boundaries and sulfur-induced copper embrittlement, copper alloys are designed in this study. Alloying elements that can act as chemical anchors to suppress sulfur diffusion and the formation of Cu₂S along grain boundaries are investigated based on the understanding of the microscopic mechanism of sulfur diffusion and Cu₂S precipitation along grain boundaries. Copper alloy ingots are experimentally manufactured to validate the alloying elements. Microstructural analysis using scanning electron microscopy with energy dispersive spectroscopy demonstrates that Cu₂S particles are not formed at grain boundaries but randomly distributed within grains in all the vacuum arc-melted Cu alloys (Cu-Si, Cu-Ag, and Cu-Zr). Further studies will be conducted to evaluate the mechanical and corrosion properties of the developed Cu alloys.

Copper alloy, Sulfur embrittlement, Disposal canister, Sulfide, Alloy design

Spin-off pyroprocessing technology and inert anode materials to replace the conventional carbon-based smelting process for critical materials were introduced. Efforts to select inert anode materials through numerical analysis and selected experimental results were devised for the high-throughput reduction of oxide feedstocks. The electrochemical properties of the inert anode material were evaluated, and stable electrolysis behavior and CaCu generation were observed during molten salt recycling. Thereafter, CuTi was prepared by reacting rutile (TiO₂) with CaCu in a Ti crucible. The formation of CuTi was confirmed when the concentration of CaO in the molten salt was controlled at 7.5mol%. A laboratory-scale electrorefining study was conducted using CuTi(Zr, Hf) alloys as the anodes, with a Ti electrodeposit conforming to the ASTM B299 standard recovered using a pilot-scale electrorefining device.

Pyroprocessing, Electroreduction, Electrorefining, Eco-friendly process, High-purity titanium, Molten salt

Coupled thermo-hydraulic-mechanical (THM) processes are essential for the long-term performance of deep geological disposal of high-level radioactive waste. In this study, a numerical sensitivity analysis was performed to analyze the effect of rock properties on THM responses after the execution of the heater test at the Kamaishi mine in Japan. The TOUGHFLAC simulator was applied for the numerical simulation assuming a continuum model for coupled THM analysis. The rock properties included in the sensitivity study were the Young’s modulus, permeability, thermal conductivity, and thermal expansion coefficients of crystalline rock, rock salt, and clay. The responses, i.e., temperature, water content, displacement, and stress, were measured at monitoring points in the buffer and near-field rock mass during the simulations. The thermal conductivity had an overarching impact on THM responses. The influence of Young’s modulus was evident in the mechanical behavior, whereas that of permeability was noticed through the change in the temperature and water content. The difference in the THM responses of the three rock type models implies the importance of the appropriate characterization of rock mass properties with regard to the performance assessment of the deep geological disposal of high-level radioactive waste.

Deep geological disposal, Coupled THM process, Near-field rock mass, Kamaishi mine experiment

Sulfide concentrations critically affect worker safety and the integrities of underground facilities, such as deep geological repositories for spent nuclear fuel. Sulfide is highly sensitive to oxygen, which can oxidize sulfide to sulfate. This can hinder precise measurement of the sulfide concentration. Hence, a literature review was conducted, which revealed that two methods are commonly used: the methylene blue and sulfide ion-selective electrode (ISE) methods. Inductively coupled plasma optical emission spectroscopy (ICP-OES) was used for comparison with the two methods. The sulfide ISE method was found to be superior as it yielded results with a higher degree of accuracy and involved fewer procedures for quantification of the sulfide concentration in solution. ICP-OES results can be distorted significantly when sulfide is present in solution owing to the formation of H₂S gas in the ICP-OES nebulizer. Therefore, the ICP-OES must be used with caution when quantifying underground water to prevent any distortion in the measured results. The results also suggest important measures to avoid problems when using ICP-OES for site selection. Furthermore, the sulfide ISE method is useful in determining sulfide concentrations in the field to predict the lifetime of disposal canisters of spent nuclear fuel in deep geological repositories and other industries.

Sulfide, Ion-selective electrode, Deep geological repository, Titration, Methylene blue, Inductively coupled plasma optical emission spectroscopy

There is a gap in our understanding of the behavior of fused and molten fuel salts containing unavoidable contamination, such as those due to fabrication, handling, or storage. Therefore, this work used calorimetry to investigate the change in liquidus temperature of PuCl₃, having an unknown purity and that had been in storage for several decades. Further research was performed by additions of NaCl, making several compositions within the binary system, and summarizing the resulting changes, if any, to the phase diagram. The melting temperature of the PuCl₃ was determined to be 746.5°C, approximately 20°C lower than literature reported values, most likely due to an excess of Pu metal in the PuCl₃ either due to the presence of metallic plutonium remaining from incomplete chlorination or due to the solubility of Pu in PuCl₃. From the melting temperature, it was determined that the PuCl₃ contained between 5.9 to 6.2mol% Pu metal. Analysis of the NaCl-PuCl₃ samples showed that using the Pu rich PuCl₃ resulted in significant changes to the NaCl-PuCl₃ phase diagram. Most notably an unreported phase transition occurring at approximately 406°C and a new eutectic composition of 52.7mol% NaCl–38.7mol% PuCl₃–2.5mol% Pu which melted at 449.3°C. Additionally, an increase in the liquidus temperatures was seen for NaCl rich compositions while lower liquidus temperatures were seen for PuCl₃ rich compositions. It can therefore be concluded that changes will occur in the NaCl-PuCl₃ binary system when using PuCl₃ with excess Pu metal. However, melting temperature analysis can provide valuable insight into the composition of the PuCl3 and therefore the NaCl-PuCl3 system.

Molten salt, Plutonium trichloride, Phase diagram, Melting temperature, Calorimetry

Several countries have been operating radioactive waste disposal (RWD) programs to construct their own repositories and have used natural analogues (NA) studies directly or indirectly to ensure the reliability of the long-term safety of deep geological disposal (DGD) systems. A DGD system in Korea has been under development, and for this purpose a generic NA study is necessary. The Korea Atomic Energy Research Institute has just launched the first national NA R&D program in Korea to identify the role of NA studies and to support the safety case in the RWD program. In this article, we review some cases of NA studies carried out in advanced countries considering crystalline rocks as candidate host rocks for high-level radioactive waste disposal. We examine the differences among these case studies and their roles in reflecting each country’s disposal repository design. The legal basis and roadmap for NA studies in each country are also described. However because the results of this analysis depend upon different environmental conditions, they can be only used as important data for establishing various research strategies to strengthen the NA study environment for domestic disposal system research in Korea.

Deep geological disposal, Disposal system, Natural analogue study, Radioactive waste disposal program, Natural analogue programv

During the decades after the Fukushima Daiichi Nuclear Power Station (FDNPS) accident, ambient dose rates have markedly decreased when compared to those at the early state of the accident. Government projects have been continuously conducted by surveying the ambient dose rate and radiocesium distributions. Airborne surveys using crewed helicopters and unmanned aerial vehicles (UAVs) are the best methods for obtaining an overall picture of the distribution. However, ground-based surveys are required for accurate measurements near the population. The differences between these methods include the knowledge of the post depositional behavior of radionuclides in land use. The survey results form the basis for policy decisions such as lifting evacuation zones, decontamination, and other countermeasures. These surveys contain crucial findings regarding post-accident responses. This paper reviews the survey methods of government projects and current situation around the FDNPS. The visualization methods and databases of ambient dose rates are also reviewed to provide information to the population.

Fukushima Daiichi Nuclear Power Station accident, Ambient dose rate, Airborne radiation survey, Unmanned aerial vehicles

The Fukushima accident in 2011 revealed some major flaws in traditional nuclear fuel materials under accidental conditions. Thus, the focus of research has shifted toward “accident tolerant fuel” (ATF). The aim of this approach is to develop fuel material solutions that lead to improved reactor safety. The application of protective coatings on the surface of nuclear fuel cladding has been proposed as a near-term solution within the ATF framework. Many coating materials are being developed and evaluated. In this article, an overview of different zirconium-based alloys currently in use in the nuclear industry is provided, and their performances in normal and accidental conditions are discussed. Coating materials proposed by different institutions and organizations, their performances under different conditions simulating nuclear reactor environments are reviewed. The strengths and weaknesses of these coatings are highlighted, and the challenges addressed by different studies are summarized, providing a basis for future research. Finally, technologies and methods used to synthesize thin-film coatings are outlined.

Accident tolerant fuel, Protective coatings, Nuclear fuel cladding, Corrosion of Zr-based alloys

As part of the safety case development for generic disposal sites in Korea, it is necessary to develop generic assessment models using various geosphere–biosphere interfaces (GBIs) and potentially exposed groups (PEGs) that reflect the natural environmental characteristics and the lifestyles of people in Korea. In this study, a unique modeling strategy was developed to systematically construct and select Korean generic biosphere assessment models. The strategy includes three process steps (combination, screening, and experts’ scoring) for the biosphere system conditions. First, various conditions, such as climate, topography, GBIs, and PEGs, were combined in the biosphere system. Second, the combined calculation cases were configured into interrelation matrices to screen out some calculation cases that were highly unlikely or less significant in terms of the exposure dose. Finally, the selected calculation cases were prioritized based on expert judgment by scoring the knowledge, probability, and importance. The results of this study can be implemented in the development of biosphere assessment models for Korean generic sites. It is believed that this systematic methodology for selecting the candidate calculation cases can contribute to increasing the confidence of future site-specific biosphere assessment models.

Deep geological disposal, Biosphere assessment, Generic modeling strategy, Climate change & Landscape development, Biosphere system conditions

Various linear system solvers with multi-physics analysis schemes are compared focusing on the near-field region considering thermal-hydraulic-chemical (THC) coupled multi-physics phenomena. APro, developed at KAERI for total system performance assessment (TSPA), performs a finite element analysis with COMSOL, for which the various combinations of linear system solvers and multi-physics analysis schemes should to be compared. The KBS-3 type disposal system proposed by Sweden is set as the target system and the near-field region, which accounts for most of the computational burden is considered. For comparison of numerical analysis methods, the computing time and memory requirement are the main concerns and thus the simulation time is set up to one year. With a single deposition hole problem, PARDISO and GMRESSSOR are selected as representative direct and iterative solvers respectively. The performance of representative linear system solvers is then examined through a problem with an increasing number of deposition holes and the GMRES-SSOR solver with a segregated scheme shows the best performance with respect to the computing time and memory requirement. The results of the comparative analysis are expected to provide a good guideline to choose better numerical analysis methods for TSPA.

Linear system solver, Multi-physics analysis scheme, Total system performance assessment, Apro

APro, a modularized process-based total system performance assessment framework, was developed at the Korea Atomic Energy Research Institute (KAERI) to simulate radionuclide transport considering coupled thermal-hydraulic-mechanicalchemical processes occurring in a geological disposal system. For reactive transport simulation considering geochemical reactions, COMSOL and PHREEQC are coupled with MATLAB in APro using an operator splitting scheme. Conventionally, coupling is performed within a MATLAB interface so that COMSOL stops the calculation to deliver the solution to PHREEQC and restarts to continue the simulation after receiving the solution from PHREEQC at every time step. This is inefficient when the solution is frequently interchanged because restarting the simulation in COMSOL requires an unnecessary setup process. To overcome this issue, a coupling scheme that calls PHREEQC inside COMSOL was developed. In this technique, PHREEQC is called through the “MATLAB function” feature, and PHREEQC results are updated using the COMSOL “Pointwise Constraint” feature. For the one-dimensional advection-reaction-dispersion problem, the proposed coupling technique was verified by comparison with the conventional coupling technique, and it improved the computation time for all test cases. Specifically, the more frequent the link between COMSOL and PHREEQC, the more pronounced was the performance improvement using the proposed technique.

COMSOL, PHREEQC, MATLAB, Operator splitting, SNIA

The potential impact of hypothetical accidents that occur during the immediate and deferred dismantling of the Kori Unit 1 steam generator has been comprehensively evaluated. The evaluation includes determining the inventory of radionuclides in the Steam Generator based on surface contamination measurements, assuming a rate of release for each accident scenario, and applying external and internal exposure dose coefficients to assess the effects of radionuclides on human health. The evaluation also includes calculating the atmospheric dispersion factor using the PAVAN code and analyzing three years of meteorological data from Kori NPP to determine the degree of diffusion of radionuclides in the atmosphere. Overall, the effective dose for residents living in the Exclusion Area Boundary (EAB) of Kori NPP is predicted, an it is found that the maximum level of the dose is 0.034% compared to the annual dose limit of 1 mSv for the general public. This implies that the potential impact of hypothetical accidents on human health discussed above is within acceptable limits.

Dismantling, Steam generator, Hypothetical accidents, Atmospheric dispersion factor

To ensure radiological safety margin in the transport and storage of spent nuclear fuel, it is crucial to perform source term and shielding analyses in advance from the perspective of conservation. When performing source term analysis on UO₂ fuel, which is mostly used in commercial nuclear power plants, uranium and oxygen are basically considered to be the initial materials of the new fuel. However, the presence of impurities in the fuel and structural materials of the fuel assembly may influence the source term and shielding analyses. The impurities could be radioactive materials or the stable materials that are activated by irradiation during reactor power operation. As measuring the impurity concentration levels in the fuel and structural materials can be challenging, publicly available information on impurity concentration levels is used as a reference in this evaluation. To assess the effect of impurities, the results of the source term and shielding analyses were compared depending on whether the assumed impurity concentration is considered. For the shielding analysis, generic cask design data developed by KEPCO-E&C was utilized.

Spent nuclear fuel, Impurities, Source term, Shielding, Cask

The emission of off-gas streams from used fuel recycling is a concern in nuclear energy usage as they contain radioactive compounds, such as, ³H, ¹⁴C, ⁸⁵Kr, ¹³¹I, and ¹²⁹I that can be harmful to human health and environment. Radioactive iodine, ¹²⁹I, is particularly troublesome as it has a half-life of more than 15 million years and is prone to accumulate in human thyroid glands. Organic iodides are hazardous even at very low concentrations, and hence the capture of ¹²⁹I is extremely important. Dynamic adsorption experiments were conducted to determine the efficiency of sodium mordenite, partially exchanged silver mordenite, and fully exchanged silver mordenite for the removal of methyl iodide present at parts per billion concentrations in a simulated off-gas stream. Kinetic analysis of the system was conducted incorporating the effects of diffusion and mass transfer. The possible reaction mechanism is postulated and the order of the reaction and the values of the rate constants were determined from the experimental data. Adsorbent characterization is performed to investigate the nature of the adsorbent before and after iodine loading. This paper will offer a comprehensive understanding of the methyl iodide behavior when in contact with the mordenites.

¹²⁹I, Mordenite, Methyl iodide, Dynamic adsorption

In the design of HLW repositories, it is important to confirm the performance and safety of buffer materials at high temperatures. Most existing models for predicting hydraulic conductivity of bentonite buffer materials have been derived using the results of tests conducted below 100°C. However, they cannot be applied to temperatures above 100°C. This study suggests a prediction model for the hydraulic conductivity of bentonite buffer materials, valid at temperatures between 100°C and 125°C, based on different test results and values reported in literature. Among several factors, dry density and temperature were the most relevant to hydraulic conductivity and were used as important independent variables for the prediction model. The effect of temperature, which positively correlates with hydraulic conductivity, was greater than that of dry density, which negatively correlates with hydraulic conductivity. Finally, to enhance the prediction accuracy, a new parameter reflecting the effect of dry density and temperature was proposed and included in the final prediction model. Compared to the existing model, the predicted result of the final suggested model was closer to the measured values.

Hydraulic conductivity, Bentonite buffer materials, Prediction model, Engineered barrier system, Dry density and temperature

Neutron resonance transmission technique was applied for assaying isotopic fissile materials produced in the pyro-process. In each process of the pyro-process, a different composition of the fissile material is produced. Simulation was basically performed on ²³⁵U and ²³⁹Pu assay for TRU-RE product, hull waste, and uranium addition. The resonance energies were evaluated for uranium and plutonium in the simulation, and the linearity in the detection response was examined on the fissile content variation. The linear resonance energies were determined for the analysis of ²³⁵U and ²³⁹Pu on the different fissile materials. For enriched TRU-RE assay, the sample condition was suggested; The sample density, content, and thickness are the key factors to obtain accurate fissile content. The detection signal is discriminated for uranium and plutonium in neutron resonance technique. The transmitted signal for fissile resonance has a direct relation with the content of fissile. The simulation results indicated that the neutron resonance technique is promising to analyze 235U and 239Pu for various types of the pyro-process material. An accurate fissile assay will contribute toward safeguarding the pyro-processing system.

Pyro-process, Fissile analysis, Neutron resonance, Measurement, Accountancy

According to NSSC Notice No. 2021-10, safety analysis needs to be introduced in the decommissioning plan. Public and occupational dose analyses should be conducted, specifically for unexpected radiological accidents. Herein, based on the risk matrix and analytic hierarchy process, the method of selecting accident scenarios during the decommissioning of nuclear power plants has been proposed. During decommissioning, the generated spent resin exhibits relatively higher activity than other generated wastes. When accidents occur, the release fraction varies depending on the conditioning method of radioactive waste and type of radioactive nuclides or accidents. Occupational dose analyses for 2 (fire and drop) among 11 accident scenarios have been performed. The radiation doses of the additional exposures caused by the fire and drop accidents are 1.67 and 4.77 mSv, respectively.

Occupational dose analysis, Decommissioning, Fire accident, Drop accident, AHP method

Compared to operational wastes, nuclear power plant (NPP) decommissioning wastes are generated in larger quantities within a short time and include diverse types with a wider range of radiation characteristics. Currently used 200 L drums and IP-2 type transport containers are inefficient and restrictive in packaging and transporting decommissioning wastes. Therefore, new packaging and transport containers with greater size, loading weight, and shielding performance have been developed. When transporting radioactive materials, radiological safety should be assessed by reflecting parameters such as the type and quantity of the package, transport route, and transport environment. Thus far, safety evaluations of radioactive waste transport have mainly targeted operational wastes, that have less radioactivity and a smaller amount per transport than decommissioning wastes. Therefore, in this study, the possible radiation effects during the transport from NPP to disposal facilities were evaluated to reflect the characteristics of the newly developed containers and decommissioning wastes. According to the evaluation results, the exposure dose to transport workers, handling workers, and the public was lower than the domestic regulatory limit. In addition, all exposure dose results were confirmed, through sensitivity analysis, to satisfy the evaluation criteria even under circumstances when radioactive materials were released 100% from the container.

Decommissioning radioactive waste, RADTRAN, Transport risk assessment, LILW disposal facility, Annual dose limit

A study was conducted on the vitrification of the rare earth oxide waste generated from the PyroGreen process. The target rare earth waste consisted of eight elements: Nd, Ce, La, Pr, Sm, Y, Gd, and Eu. The waste loading of the rare earth waste in the developed borosilicate glass system was 20wt%. The fabricated glass, processed at 1,200℃, exhibited uniform and homogeneous surface without any crystallization and precipitation. The viscosity and electrical conductivity of the melted glass at 1,200℃ were 7.2 poise and 1.1 S·cm⁻¹, respectively, that were suitable for the operation of the vitrification facility. The calculated leaching index of Cs, Co, and Sr were 10.4, 10.6, and 9.8, respectively. The evaluated Product Consistency Test (PCT) normalized release of the glass indicated that the glass satisfied the requirements for the disposal acceptance criteria. Furthermore, the pristine, 90 days water immersed, 30 thermal cycled, and 10 MGy gamma ray irradiated glasses exhibited good compressive strength. The results indicated that the fabricated glass containing rare earth waste from the PyroGreen process was acceptable for the disposal in the repository, in terms of chemical durability and mechanical strength.

Vitrification, Rare earth oxide, Characteristics, PCT, Borosilicate

APro, developed in KAERI for the process-based total system performance assessment (TSPA) of deep geological disposal systems, performs finite element method (FEM)-based multiphysics analysis. In the FEM-based analysis, the mesh element quality influences the numerical solution accuracy, memory requirement, and computation time. Therefore, an appropriate mesh structure should be constructed before the mesh stability analysis to achieve an accurate and efficient process-based TSPA. A generic reference case of DECOVALEX-2023 Task F, which has been proposed for simulating stationary groundwater flow and time-dependent conservative transport of two tracers, was used in this study for mesh stability analysis. The relative differences in tracer concentration varying mesh structures were determined by comparing with the results for the finest mesh structure. For calculation efficiency, the memory requirements and computation time were compared. Based on the mesh stability analysis, an approach based on adaptive mesh refinement was developed to resolve the error in the early stage of the simulation time-period. It was observed that the relative difference in the tracer concentration significantly decreased with high calculation efficiency.

Mesh stability, DECOVALEX-2023 Task F, Adaptive mesh refinement, Total system performance assessment, APro

To broaden the utilization of nuclear energy, uranium as a fuel should be mined indispensably. Mining accounts for the largest portion of the cost of producing the uranium assembly. Therefore, this study analyzes the trends of uranium prices, which have a significant impacts on the mining cost. Uranium production contributing to the price fluctuations is explained in five periods from 1945 to the present. Moreover, the series of events affecting uranium prices from the 1970s until the present are verified. Among them, the most recent incidents considered in this study are the following: COVID-19 pandemic, Kazakhstan unrest, and Russia-Ukraine war. European countries have started to reconsider the transition to nuclear power to reduce their dependence on Russian oil and gas, which has contributed to the surge in uranium prices. Based on the results of this study, various international issues have been closely associated with the nuclear power industry and uranium, affecting the production of uranium and its price.

Uranium oxide (U₃O₈), Uranium production, Uranium price

Spent fuels (SFs) are stored in a storage pool after discharge from nuclear power plants. They can be transferred to for the further processes such as dry storage sites, processing plants, or disposal sites. One of important measures of SF is the burnup. Since the radioactivity of SF is strongly dependent on its burnup, the burnup of SF should be well estimated for the safe management, storage, and final disposal. Published papers about the methodology for the burnup estimation from the known activities of important radioactive sources are somewhat rare. In this study, we analyzed the dependency of the burnup on the important radiation source activities using ORIGEN-ARP, and suggested simple correlations that relate the burnup and the important source activities directly. A burnup estimation equation is suggested for PWR fuels relating burnup with total neutron source intensity (TNSI), initial enrichment, and cooling time. And three burnup estimation equations for major gamma sources, ¹³⁷Cs, ¹³⁴Cs, and ¹⁵⁴Eu are also suggested.

Spent nuclear fuel, Neutron and gamma sources, Burnup estimation, Enrichment, Cooling time

In South Korea, the exporters of items related to nuclear power generation are diversified. Consequently, there is a risk of illegitimate export by companies failing to recognize the export control system because the awareness about this system for the strategic items among the subcontractors of nuclear power facilities is limited. To prevent illegitimate export of the strategic items, it is necessary to conduct outreach activities regarding the export control system for the related companies. Additionally, the exporters and export license examiners should consider whether an export target is on the Denial List, who may divert the strategic items to weapons of mass destruction. Therefore, the Korea Institute of Nuclear Nonproliferation and Control developed two systems for controlling illegitimate export of the Trigger List items. The first system, Nuclear Industry Information Collection and Analysis System, can gather information about the key nuclear industries in Korea and analyze the dealing of strategic items. The second system, Denied Persons Information Gathering System, can regularly gather information about the denied persons and provide the updated data to the exporters and regulatory examiners. These two systems can be used for outreach activities and export license examination to prevent illegitimate export of the strategic items.

Illegitimate export, Strategic items, Information collection, Denial list, the Export control

This paper described a method for analyzing the structural performance of a metal container used for disposing radioactive waste generated during the decommissioning of a nuclear power plant, and numerical analysis results of a method for reinforcing the container. The containers to be analyzed were those that can be used in near-surface and landfill disposal facilities scheduled to be operated at the Gyeongju radioactive waste disposal facility. Structural reinforcement of the container was performed by lattice reinforcement, column reinforcement, and bottom plate reinforcement. Accordingly, a total of 14 reinforcement cases were modeled. The external force causing damage to the container was set equivalent to the impact of a 9-m fall, accounting for the height of the vault at the near-surface disposal facility. The reinforcement methods with a high contribution to the structural performance of the container were concluded to be lattice and column reinforcements.

NPP decommissioning radioactive waste, Near-surface disposal facility, Disposal container, FEM analysis

United States Nuclear Regulatory Commission (U.S. NRC) specifies regulations on obtaining licenses and describes the technical position on the average waste concentration, also known as Concentration Averaging and Encapsulation Branch Technical Position (CA BTP); CA BTP helps classify blendable waste and discrete items and address concentration averaging. The technical position details are reviewed and compared in a real environment in Korea. A few cases of concentration averaging based on the application of CA BTP to domestic radioactive waste are presented, and the feasibility of the application is assessed. The radioactive waste considered herein does not satisfy the Disposal Concentration Limit (DCL) of the second-phase disposal facility while applying the preliminary classification. However, if CA BTP is applied when the radioactive waste is mixed with other radioactive waste items in a large and heavy container, it can be disposed of at the second-phase disposal facility in Gyeongju Repository. To apply the CA BTP of the U.S. NRC, it is necessary to investigate the safety assessment conditions of the US and Korea.

CA BTP, Blendable waste, Discrete item, Disposal Concentration Limit, 2nd phase disposal facility

The thermal integrity of spent nuclear fuels has to be maintained during their long-term dry storage. The detailed temperature distributions of spent fuel assemblies are essential for evaluating the integrity of their dry storage systems. In this study, a subchannel analysis model was developed for a canister of a single fuel assembly using the COBRA-SFS code. The thermal parameters affecting the peak cladding temperature (PCT) of the spent fuel assembly were identified, and sensitivity analyses were performed based on these parameters. The subchannel analysis results indicated the presence of a recirculation flow, based on natural convection, between the fuel assembly and downcomer region. The sensitivity analysis of the thermal parameters indicated that the PCT was affected by the emissivity of the fuel cladding and basket, convective heat transfer coefficient, and thermal conductivity of the fluid. However, the effects of the wall friction factor of the canister, form loss coefficient of the grid spacers, and thermal conductivities of the solid materials, on the PCT were predominantly ignored.

Spent nuclear fuel, Dry storage, Peak cladding temperature, Subchannel analysis, Sensitivity analysis, Convective heat transfer

Evaluating the effectiveness of the radiation protection measures deployed at the Centralized Radioactive Waste Management Facility in Ghana is pivotal to guaranteeing the safety of personnel, public and the environment, thus the need for this study. Radiagem^TM 2000 was used in measuring the dose rate of the facility whilst the personal radiation exposure of the personnel from 2011 to 2022 was measured from the thermoluminescent dosimeter badges using Harshaw 6600 Plus Automated TLD Reader. The decay store containing scrap metals from dismantled disused sealed radioactive sources (DSRS), and low-level wastes measured the highest dose rate of 1.06 ± 0.92 µSv·h⁻¹. The range of the mean annual average personnel dose equivalent is 0.41–2.07 mSv. The annual effective doses are below the ICRP limit of 20 mSv. From the multivariate principal component analysis biplot, all the personal dose equivalent formed a cluster, and the cluster is mostly influenced by the radiological data from the outer wall surface of the facility where no DSRS are stored. The personal dose equivalents are not primarily due to the radiation exposures of staff during operations with DSRS at the facility but can be attributed to environmental radiation, thus the current radiation protection measures at the Facility can be deemed as effective.

Radioactive waste, Disused sealed radioactive sources, Multivariate analysis, Principal component analysis, Radiological data, Thermoluminescent dosimeter

Since 1997, the Republic of Korea (ROK) has been developing pyro-processing (Pyro) technology to reduce the disposal burden of high-level radioactive waste by recycling spent nuclear fuel (SNF). Compared to plutonium and uranium extraction process, Korean Pyro technology has relatively excellent proliferation resistance that cannot separate pure plutonium owing to its intrinsic characteristics. Regarding Pyro technology development of ROK, the Bush administration considered that Pyro is not reprocessing under the Global Nuclear Energy Partnership, whereas the Obama administration considered that Pyro is subject to reprocessing. However, the Bush and Obama administrations did not allow ROK to conduct full Pyro activities using SNF, even though ROK had faithfully complied with international nonproliferation obligations. This is because the US nuclear nonproliferation policy to prevent the spread of sensitive technologies, such as enrichment and reprocessing, has a strong effect on ROK, unlike Japan, on a bilateral level beyond the NPT regime for non-proliferation of nuclear weapons.

Pyroprocessing, Nuclear, Nonproliferation, Reprocessing, GNEP, Nuclear cooperation agreement

This study examined the efficacy of various chlorinating agents in partitioning light water reactor spent fuel, with the aim of optimizing the chlorination process. Through thermodynamic equilibrium calculations, we assessed the outcomes of employing MgCl₂, NH₄Cl, and Cl₂ as chlorinating agents. A comparison was drawn between using a single agent and a sequential approach involving all three agents (MgCl₂, NH₄Cl, and Cl₂). Following heat treatment, the utilization of MgCl₂ as the sole chlorinating agent resulted in a moderate separation. Specifically, this method yielded a solid separation with 96.9% mass retention, 31.7% radioactivity, and 44.2% decay heat, relative to the initial spent fuel. In contrast, the sequential application of the chlorinating agents following heat treatment led to a final solid separation characterized by 93.1% mass retention, 5.1% radioactivity, and 15.4% decay heat, relative to the original spent fuel. The findings underscore the potential effectiveness of a sequential chlorination strategy for partitioning spent fuel. This approach holds promise as a standalone technique or as a complementary process alongside other partitioning processes such as pyroprocessing. Overall, our findings contribute to the advancement of spent fuel management strategies.

Thermodynamic calculation, Equilibrium composition, Sequential chlorination, Spent fuel, Partitioning

During the operation of a nuclear power plant (NPP), the generation of radioactive waste, including dry active waste (DAW), concentrates, spent resin, and filters, mandates the implementation of appropriate disposal methods to adhere to Korea’s waste acceptance criteria (WAC). In this context, this study investigates the potential use of polymer concrete (PC) as a high-integrity container (HIC) material for solidifying and packaging these waste materials. PC is a versatile composite material comprising binding polymers, aggregates, and additives, known for its exceptional strength and chemical stability. A comprehensive analysis of PC’s long-term integrity was conducted in this study. First, its compressive strength, which is crucial for ensuring the structural stability of HICs over extended periods, was evaluated. Subsequently, the resilience of PC was tested under various stress conditions, including biological, radiological, thermal, and chemical stressors. The findings of this study indicate that PC exhibits remarkable long-term properties, demonstrating exceptional stability even when subjected to diverse stressors. The results therefore underscore the potential viability of PC as a reliable material for constructing high-integrity containers, thus contributing to the safe and sustainable management of radioactive waste in NPPs.

Polymer concrete, High integrity container, Long term stability

The Radiation and Decommissioning Laboratory of Central Research Institute (CRI) of Korea Hydro and Nuclear Power Co. (KHNP) performs research to technically support the effective management of radiological hazards to avoid risks to civilians, the workers, and the environment from the radiological risks. The laboratory mainly consists of three technical groups: decommissioning and SF technology group, radiation and chemistry group, and radwaste and environment group. The groups carry out various R&D such as decommissioning, spent fuel management, radiation protection, water chemistry management, and radioactive waste management. The laboratory also technically supports the calibration of radiometric instruments as a Korea Laboratory Accreditation Scheme (KOLAS), approval for decommissioning, guidance for radioactive waste management, state-of-the-art technology evaluations, and technology transfer.

Radiation and decommissioning laboratory, Decommissioning, Spent fuel, Water chemistry, Radioactive waste

Thermodynamic sorption modeling can enhance confidence in assessing and demonstrating the radionuclide sorption phenomena onto various mineral adsorbents. In this work, Ca-montmorillonite was successfully purified from Bentonil-WRK bentonite by performing the sequential physical and chemical treatments, and its geochemical properties were characterized using X-ray diffraction, Brunauer-Emmett-Teller analysis, cesium-saturation method, and controlled continuous acidbase titration. Further, batch experiments were conducted to evaluate the adsorption properties of Cs(I) and Sr(II) onto the homoionic Ca-montmorillonite under ambient conditions, and the diffuse double layer model-based inverse analysis of sorption data was performed to establish the relevant surface reaction models and obtain corresponding thermodynamic constants. Two types of surface reactions were identified as responsible for the sorption of Cs(I) and Sr(II) onto Ca-montmorillonite: cation exchange at interlayer site and complexation with edge silanol functionality. The thermodynamic sorption modeling provides acceptable representations of the experimental data, and the species distributions calculated using the resulting reaction constants accounts for the predominance of cation exchange mechanism of Cs(I) and Sr(II) under the ambient aqueous conditions. The surface complexation of cationic fission products with silanol group slightly facilitates their sorption at pH > 8.

Thermodynamic sorption modeling, Bentonil-WRK montmorillonite, Cesium, Strontium, Cation exchange, Surface complexation

In this study, the impact load resulting from collision with the fuel rods of surrogate spent nuclear fuel (SNF) assemblies was measured during a rolling test based on an analysis of the data from surrogate SNF-loaded sea transportation tests. Unfortunately, during the sea transportation tests, excessive rolling motion occurred on the ship during the test, causing the assemblies to slip and collide with the canister. Hence, we designed and conducted a separate test to simulate rolling in sea transportation to determine whether such impact loads can occur under normal conditions of SNF transport, with the test conditions for the fuel assembly to slide within the basket experimentally determined. Rolling tests were conducted while varying the rolling angle and frequency to determine the angles and frequencies at which the assemblies experienced slippage. The test results show that slippage of SNF assemblies can occur at angles of approximately 14° or greater because of rolling motion, which can generate impact loads. However, this result exceeds the conditions under which a vessel can depart for coastal navigation, thus deviating from the normal conditions required for SNF transport. Consequently, it is not necessary to consider such loads when evaluating the integrity of SNFs under normal transportation conditions.

Normal conditions of transport, Rolling test, Sea transportation test, Spent nuclear fuel, Transportability

With South Korea increasingly focusing on nuclear energy, the management of spent nuclear fuel has attracted considerable attention in South Korea. This study established a novel procedure for selecting safety-relevant radionuclides for long-term safety assessments of a deep geological repository in South Korea. Statistical evaluations were performed to identify the design basis reference spent nuclear fuels and evaluate the source term for up to one million years. Safety-relevant radionuclides were determined based on the half-life criteria, the projected activities for the design basis reference spent nuclear fuel, and the annual limit of ingestion set by the Nuclear Safety and Security Commission Notification No. 2019-10 without considering their chemical and hydrogeological properties. The proposed process was used to select 56 radionuclides, comprising 27 fission and activation products and 29 actinide nuclides. This study explains first the determination of the design basis reference spent nuclear fuels, followed by a comprehensive discussion on the selection criteria and methodology for safety-relevant radionuclides.

Deep geological repository, Long-term safety assessment, Safety-relevant radionuclide, Design basis reference spent nuclear fuel, Source terms, SCALE 6.2.4

To ensure the safety of disposal facilities for radioactive waste, it is essential to quantitatively evaluate the performance of the waste disposal facilities by using safety assessment models. This paper addresses the development of the safety assessment model for the underground silo of Wolseong Low-and Immediate-Level Waste (LILW) disposal facility in Korea. As the simulated result, the nuclides diffused from the waste were kept inside the silo without the leakage of those while the integrity of the concrete is maintained. After the degradation of concrete, radionuclides migrate in the same direction as the groundwater flow by mainly advection mechanism. The release of radionuclides has a positive linear relationship with a half-life in the range of medium half-life. Additionally, the solidified waste form delays and reduces the migration of radionuclides through the interaction between the nuclides and the solidified medium. Herein, the phenomenon of this delay was implemented with the mass transfer coefficient of the flux node at numerical modeling. The solidification effects, which are delaying and reducing the leakage of nuclides, were maintained the integrity of the nuclides. This effect was decreased by increasing the half-life and the mass transfer coefficient of radionuclides.

Waste disposal facility, Safety assessment model, Near-field, Nuclide release

Concerning the apprehensions about naturally occurring radioactive materials (NORM) residues, the International Atomic Energy Agency (IAEA) and its member nations have acknowledged the imperative to ensure the radiation safety of NORM industries. Residues with elevated radioactivity concentrations are predominantly produced during NORM processing, in the form of scale and sludge, referred to as technically enhanced NORM (TENORM). Substantial quantities of TENORM residues have been released externally due to the dismantling of NORM processing factories. These residues become concentrated and fixed in scale inside scrap pipes. To assess the radioactivity of scales in pipes of various shapes, a Monte Carlo simulation was employed to determine dose rates corresponding to the action level in TENORM regulations for different pipe diameters and thicknesses. Onsite gamma spectrometry was conducted on a scrap iron pipe from the titanium dioxide manufacturing factory. The measured dose rate on the pipe enabled the estimation of NORM concentration in the pipe scale onsite. The derived action level in dose rate can be applied in the NORM regulation procedure for on-site judgments.

NORM, TENORM, NORM residue, Residual radioactivity, In situ measurement

This article presents the crucial role played by the French underground research laboratory (URL) in initiating the deep geological repository project Cigéo. In January 2023, Andra finalized the license application for the initial construction of Cigéo. Depending on Government’s decision, the construction of Cigéo may be authorized around 2027. Cigéo is the result of a National program, launched in 1991, aiming to safely manage high-level and intermediate level long-lived radioactive wastes. This National program is based on four principles: 1) excellent science and technical knowledge, 2) safety and security as primary goals for waste management, 3) high requirements for environment protection, 4) transparent and openpublic exchanges preceding the democratic decisions and orientations by the Parliament. The research and development (R&D) activities carried out in the URL supported the design and the safety demonstration of the Cigéo project. Moreover, running the URL has provided an opportunity to gain practical experience with regard to the security of underground operations, assessment of environmental impacts, and involvement of the public in the preparation of decisions. The practices implemented have helped gradually build confidence in the Cigéo project.

High-level radioactive waste, Intermediate level radioactive waste long-lived, Deep geological disposal, Cigéo, URL, Siting process

A transfer cask serves as the container for transporting and handling canisters loaded with spent nuclear fuels from light water reactors. This study focuses on a cylindrical transfer cask, standing at 5,300 mm with an external diameter of 2,170 mm, featuring impact limiters on the top and bottom sides. The base of the cask body has an openable/closable lid for loading canisters with storage modules. The transfer cask houses a canister containing spent nuclear fuels from lightweight reactors, serving as the confinement boundary while the cask itself lacks the confinement structure. The objective of this study was to conduct a structural analysis evaluation of the transfer cask, currently under development in Korea, ensuring its safety. This evaluation encompasses analyses of loads under normal, off-normal, and accident conditions, adhering to NUREG-2215. Structural integrity was assessed by comparing combined results for each load against stress limits. The results confirm that the transfer cask meets stress limits across normal, off-normal, and accident conditions, establishing its structural safety.

Transfer cask, Spent nuclear fuel, Dry storage module, Canister, Confinement boundary

The objectives of this paper are: (1) to conduct the thermal analyses of the disposal cell using COMSOL Multiphysics; (2) to determine whether the design of the disposal cell satisfies the thermal design requirement; and (3) to evaluate the effect of design modifications on the temperature of the disposal cell. Specifically, the analysis incorporated a heterogeneous model of 236 fuel rod heat sources of spent nuclear fuel (SNF) to improve the reality of the modeling. In the reference case, the design, featuring 8 m between deposition holes and 30 m between deposition tunnels for 40 years of the SNF cooling time, did not meet the design requirement. For the first modified case, the designs with 9 m and 10 m between the deposition holes for the cooling time of 40 years and five spacings for 50 and 60 years were found to meet the requirement. For the second modified case, the designs with 35 m and 40 m between the deposition tunnels for 40 years, 25 m to 40 m for 50 years and five spacings for 60 years also met the requirement. This study contributes to the advancement of the thermal analysis technique of a disposal cell.

Thermal analysis, Disposal canister, Deposition hole, Deposition tunnel, Spent nuclear fuel, COMSOL Multiphysics

The Wolsong Nuclear Power Plant (NPP) operates an on-site spent fuel dry storage facility using concrete silo and vertical module systems. This facility must be safely maintained until the spent nuclear fuel (SNF) is transferred to an external interim or final disposal facility, aligning with national policies on spent nuclear fuel management. The concrete silo system, operational since 1992, requires an aging management review for its long-term operation and potential license renewal. This involves comparing aging management programs of different dry storage systems against the U.S. NRC’s guidelines for license renewal of spent nuclear fuel dry storage facilities and the U.S. DOE’s program for long-term storage. Based on this comparison, a specific aging management program for the silo system was developed. Furthermore, the facility’s current practices—periodic checks of surface dose rate, contamination, weld integrity, leakage, surface and groundwater, cumulative dose, and concrete structure—were evaluated for their suitability in managing the silo system’s aging. Based on this review, several improvements were proposed.

Spent fuel, Dry storage system, Aging management, License renewal, Periodic verification

Currently, Japan is undertaking a nationwide project to measure and map radioactive contamination around Fukushima, as part of the efforts to restore normalcy following the nuclear accident. The Japan Atomic Energy Agency (JAEA) manages the Fukushima Environmental Safety Center, located approximately 20 km north of the Fukushima Daiichi nuclear power plant in Minamisōma City, Fukushima Prefecture. In collaboration with the JAEA, this study involved conducting comparison experiments and analyses with radiation detectors in high radiation environments, a challenging task in Korean environments. Environmental radiation surveys were conducted using three types of detectors: CZT, NaI(Tl), and LaBr₃(Ce), across two contaminated areas. Dose rate values were converted using dose rate conversion factors for each detector type, and dose rate maps were subsequently created and compared. The detectors yielded similar results, demonstrating their feasibility and reliability in high radiation environments. The findings of this study are expected to be a crucial reference for enhancing the verification and supplementation of procedures and methods in future radiation measurements and mobile surveys in high-radiation environments, using these three types of radiation instruments.

Radiation monitoring, In situ radioactivity measurement, Gamma ray spectrum, Radioactivity walking survey

The aim of this study is to ensure the structural integrity of a canister to be used in a dry storage system currently being developed in Korea. Based on burnup and cooling periods, the canister is designed with 24 bundles of spent nuclear fuel stored inside it. It is a cylindrical structure with a height of 4,890 mm, an internal diameter of 1,708 mm, and an inner length of 4,590 mm. The canister lid is fixed with multiple seals and welds to maintain its confinement boundary to prevent the leakage of radioactive waste. The canister is evaluated under different loads that may be generated under normal, off-normal, and accident conditions, and combinations of these loads are compared against the allowable stress thresholds to assess its structural integrity in accordance with NUREG-2215. The evaluation result shows that the stress intensities applied on the canister under normal, off-normal, and accident conditions are below the allowable stress thresholds, thus confirming its structural integrity.

Canister, Spent nuclear fuel, Dry storage module, Confinement boundary

The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

GIF (Generation IV international forum), Generation IV reactor, SMR (Small modular reactor), MSR (Molten salt reactor), GIF technical roadmap

Bentonite is the most probable candidate to be used as a buffer in a deep geological repository with high swelling properties, hydraulic conductivity, thermal conductivity, and radionuclide sorption ability. Among them, the radionuclide sorption ability prevents or delays the transport of radionuclides into the nearby environment when an accident occurs and the radionuclide leaks from the canister, so it needs to be strengthened in terms of long-term disposal safety. Here, we proposed a surface modification method in which some inorganic additives were added to form NaP zeolite on the surface of the bentonite yielded at Yeonil, South Korea. We confirmed that the NaP zeolite was well-formed on the bentonite surface, which also increased the sorption efficiency of Cs and Sr from groundwater conditions. Both NaP and NaX zeolite can be produced and we have demonstrated that the generation mechanism of NaX and NaP is due to the number of homogeneous/heterogeneous nucleation sites and the number of nutrients supplied from an aluminosilicate gel during the surface modification process. This study showed the potential of surface modification on bentonite to enhance the safety of deep geological radioactive waste repository by improving the radionuclide sorption ability of bentonite.

Bentonite, Surface modification, NaP zeolite, Sorption efficiency, Deep geological repository

The decommissioning of a nuclear power plant generates large amounts of radioactive waste, which is of several types. Radioactive concrete powder is classified as low-level waste, which can be disposed of in a landfill. However, its safe disposal in a landfill requires that it be immobilized by solidification using cement. Herein, a safety assessment on the disposal of solidified radioactive concrete powder waste in a conceptual landfill site is performed using RESRAD. Furthermore, sensitivity analyses of certain selected input parameters are conducted to investigate their impact on exposure doses. The exposure doses are estimated, and the relative impact of each pathway on them during the disposal of this waste is assessed. The results of this study can be used to obtain information for designing a landfill site for the safe disposal of low-level radioactive waste generated from the decommissioning of a nuclear power plant.

Decommissioning waste, Solidified waste, Landfill disposal, Safety assessment, Exposure dose

Pyroprocessing is a promising technology for managing spent nuclear fuel. The nuclear material accounting of feed material is a challenging issue in safeguarding pyroprocessing facilities. The input material in pyroprocessing is in a solid-state, unlike the solution state in an input accountability tank used in conventional wet-type reprocessing. To reduce the uncertainty of the input material accounting, a double-stage homogenization process is proposed in considering the process throughput, remote controllability, and remote maintenance of an engineering-scale pyroprocessing facility. This study tests two types of mixing equipment in the proposed double-stage homogenization process using surrogate materials. The expected heterogeneity and accounting uncertainty of Pu are calculated based on the surrogate test results. The heterogeneity of Pu was 0.584% obtained from Pressurized Water Reactor (PWR) spent fuel of 59 WGd/tU when the relative standard deviation of the mass ratio, tested from the surrogate powder, is 1%. The uncertainty of the Pu accounting can be lower than 1% when the uncertainty of the spent fuel mass charged into the first mixers is 2%, and the uncertainty of the first sampling mass is 5%.

Safeguards, Pyroprocessing, Double-stage homogenization, Heterogeneity, Nuclear material accountancy, Input material accountancy

Liquid Bi pool is a candidate electrode for an electrometallurgical process in the molten LiCl-KCl eutectic to treat the spent nuclear fuels from nuclear power plants. The electrochemical behavior of Bi³⁺ ions and the electrode reaction on liquid Bi pool were investigated with the cyclic voltammetry in an environment with or without BiCl₃ in the molten LiCl-KCl eutectic. Experimental results showed that two redox reactions of Bi³⁺ on inert W electrode and the shift of cathodic peak potentials of Li⁺ and Bi³⁺ on liquid Bi pool electrode in molten LiCl-KCl eutectic. It is confirmed that the redox reaction of lithium with respect to the liquid Bi pool electrode would occur in a wide range of potentials in molten LiCl-KCl eutectic. The obtained data will be used to design the electrometallurgical process for treating actinide and lanthanide from the spent nuclear fuels and to understand the electrochemical reactions of actinide and lanthanide at liquid Bi pool electrode in the molten LiCl-KCl eutectic.

Molten salt, Liquid Bi pool electrode, Bismuth, Electrode reaction

The aim of this review is to communicate some essential knowledge of the underlying mechanism of the corrosion of structural containment alloys during molten salt reactor operation in the context of prospective online monitoring in future MSR installations. The formation of metal halide species and the progression of their concentration in the molten salt do reflect containment corrosion, tracing the depletion of alloying metals at the alloy salt interface will assure safe conditions during reactor operation. Even though the progress of alloying metal halides concentrations in the molten salt do strongly understate actual corrosion rates, their prospective 1^st order kinetics followed by near-linearly increase is attributed to homogeneous matrix corrosion. The service life of the structural containment alloy is derived from homogeneous matrix corrosion and near-surface void formation but less so from intergranular cracking (IGC) and pitting corrosion. Online monitoring of corrosion species is of particular interest for molten chloride systems since besides the expected formation of chromium chloride species CrCl₂ and CrCl₃, other metal chloride species such as FeCl₂, FeCl₃, MoCl₂, MnCl₂ and NiCl₂ will form, depending on the selected structural alloy. The metal chloride concentrations should follow, after an incubation period of about 10,000 hours, a linear projection with a positive slope and a steady increase of < 1 ppm per day. During the incubation period, metal concentration show 1^st order kinetics and increasing linearly with time^1/2. Ideally, a linear increase reflects homogeneous matrix corrosion, while a sharp increase in the metal chloride concentration could set a warning flag for potential material failure within the projected service life, e.g. as result of intergranular cracking or pitting corrosion. Continuous monitoring of metal chloride concentrations can therefore provide direct information about the mechanism of the ongoing corrosion scenario and offer valuable information for a timely warning of prospective material failure.

Containment corrosion, Molten salt reactors, Chromium, Chloride salt systems

The nuclear legacy that remains in the United Kingdom (UK) is complex and diverse. Consisting of legacy ponds and silos, redundant reprocessing plants, research facilities, and non-standard or one-off reactor designs, the clean-up of this legacy is under the stewardship of the Nuclear Decommissioning Authority (NDA). Through a mix of prompt and delayed decommissioning strategies, the NDA has made great strides in dealing with the UK’s nuclear legacy. Fuel debris and sludge removal from the legacy ponds and silos situated at Sellafield, as part of a prompt decommissioning strategy for the site, has enabled intolerable risks to be brought under control. Reactor defueling and waste retrievals across the Magnox fleet is enabling their transition to a period of care and maintenance; accelerated through the adopted ‘Lead and Learn’ approach. Bespoke decommissioning methods implemented by the NDA have also enabled the relevant site licence companies to tackle non-standard reactor designs and one-off wastes. Such approaches have potential to influence and shape nuclear decommissioning decision making activities globally, including in Korea.

Nuclear decommissioning, Nuclear decommissioning authority, Sellafield, Magnox reactors, United Kingdom, Korea

In decommissioning a nuclear power plant, numerous concrete structures need to be demolished and decontaminated. Although concrete decontamination technologies have been developed globally, concrete cutting remains problematic due to the secondary waste production and dispersion risk from concrete scabbling. To minimize workers’ radiation exposure and secondary waste in dismantling and decontaminating concrete structures, the following conceptual designs were developed. A micro-blast type scabbling technology using explosive materials and a multi-dimensional contamination measurement and artificial intelligence (AI) mapping technology capable of identifying the contamination status of concrete surfaces. Trials revealed that this technology has several merits, including nuclide identification of more than 5 nuclides, radioactivity measurement capability of 0.1–10⁷ Bq·g⁻¹, 1.5 kg robot weight for easy handling, 10 cm robot self-running capability, 100% detonator performance, decontamination factor (DF) of 100 and 8,000 cm²·hr⁻¹ decontamination speed, better than that of TWI (7,500 cm²·hr⁻¹). Hence, the micro-blast type scabbling technology is a suitable method for concrete decontamination. As the Korean explosives industry is well developed and robot and mapping systems are supported by government research and development, this scabbling technology can efficiently aid the Korean decommissioning industry.

Micro-Blast, Decommissioning, Scabbling, Nuclear power plant

The denuclearization of North Korea was unpredictable and resulted in radical changes. Despite the skepticism and disappointment surrounding denuclearization, it is important for certain verification technologies to establish what is technically possible or practically impossible, and how reliable these technical means are. This article presents the technical hurdles in nuclear verification by systematically categorizing them into issues of correctness and completeness. Moreover, it addresses the safety and security risks during the denuclearization process, including the radiological impact on humans, environmental effects, and the illegal transfer of material, information, and technologies.

Denuclearization, Nuclear weapons, Enrichment, Reprocessing, Correctness, Completeness

This study provides technical information about the nuclear fuel handling process, which consists of various subprocesses starting from new fuel receipt to spent fuel shipment at a nuclear power plant and the design requirements of fuel handling equipment. The fuel handling system is an integrated system of equipment, tools, and procedures that allow refueling, handling and storage of fuel assemblies, which comprise the fuel handling process. The understanding and reaffirming of detailed code requirements are requested for application to the design of the fuel handling and storage facility. We reviewed the design requirements of the fuel handling equipment for its adequate cooling, prevention of criticality, its operability and maintainability, and for the prevention of fuel damage and radiological release. Furthermore, we discussed additional technical issues related to upgrading the current code requirements based on the modification of the fuel handling equipment. The suggested information provided in this paper would be beneficial to enhance the safety and the reliability of the fuel handling equipment during the handling of new and spent fuel.

Fuel handling, Fuel handling equipment, Refueling, ANSI/ANS-57.1

The structural safety of prototype transport and storage containers for very low-level radioactive liquid waste was experimentally estimated for its localization development. Transport containers for radioactive liquid waste have been researched and developed, however, there are no standardized commercial containers for very low-level radioactive waste in Korea. In this study, the structural safety of the designated IP-2 type container capable of transporting and temporarily storing large amounts of very low-level liquid waste, which is generated during the operation and decommissioning of nuclear power plants, was demonstrated. The stacking and drop tests, which were conducted to determine the structural integrity of the container, verified that there was no external leakage of the contents in spite of its structural deformation due to the drop impact. This study shows the effort required for the localization of the technology used in manufacturing transport and storage containers for very low-level radioactive liquid waste, and the additional structural reinforcement of the container in which the commercial intermediate bulk container (IBC) external frame was coupled.

Radioactive waste container, Decommissioning, Nuclear power plant, Radioactive liquid waste

The design of a wooden impact limiter equipped to a transportation cask for radioactive materials was optimized. According to International Atomic Energy Agency Safety Standards, 9 m drop tests should be performed on the transportation cask to evaluate its structural integrity in a hypothetical accident condition. For impact resistance, the size of the impact limiter should be properly determined for the impact limiter to absorb the impact energy and reduce the impact force. Therefore, the design parameters of the impact limiter were optimized to obtain a feasible optimal design. The design feasibility criteria were investigated, and several objectives were defined to obtain various design solutions. Furthermore, a probabilistic approach was introduced considering the uncertainties included in an engineering system. The uncertainty of material properties was assumed to be a random variable, and the probabilistic feasibility, based on the stochastic approach, was evaluated using reliability. Monte Carlo simulation was used to calculate the reliability to ensure a proper safety margin under the influence of uncertainties. The proposed methodology can provide a useful approach for the preliminary design of the impact limiter prior to the detailed design stage.

Transportation cask, Drop test, Impact limiter, Optimization, Material uncertainty

The sorption of Eu on MX-80 bentonite in Na–Ca–Cl solutions is investigated at a molal proton concentration (pH_m) range of 3 to 10 and an ionic strength (I) range of 0.1 to 6 m (mol·kgw⁻¹). The sorption equilibrium of Eu on MX-80 is achieved within 14 to 21 d at I = 0.1 and 6 m. The sorption distribution coefficient (K_d) values of Eu for MX-80 increase as pHm increases from 3 to 6 for all I values, and they are independent of pHm between 8 and 10 at I ≥ 0.5 m. Meanwhile, at I = 0.1 m, the K_d value at pHm = 10 is slightly lower than those at pHm = 8 and 9. The K_d values are not affected by the I values between 0.5 m and 6 m, whereas the K_d value at I = 0.1 m is greater than those at I ≥ 0.5 m, except at pH_m = 10. A two-site protolysis nonelectrostatic surface complexation and cation exchange sorption model is applied to the Eu sorption data for I ≤ 4 m, and the equilibrium constants of the sorption reactions are estimated.

Eu, MX-80 bentonite, Sorption, pH and ionic strength dependences, Sorption distribution coefficient, Two-site protolysis nonelectrostatic surface complexation and cation exchange sorption model (2SPNE SC/CE model)

The effect of oxidation time on the characteristics and mechanical properties of spent nuclear fuel cladding was investigated using Raman spectroscopy, tube rupture test, and tensile test. As oxidation time increased, the Raman peak associated with the tetragonal zirconium oxide phase diminished and merged with the Raman peak associated with the monoclinic zirconium oxide phase near 333 cm⁻¹. Additionally, the other tetragonal zirconium oxide phase peak at 380 cm⁻¹ decreased after 100 d of oxidation, whereas the zirconium monoclinic oxide peak became the dominant peak. The oxidation time had no effect on the tube rupture pressure of the oxidized zirconium alloy tube. However, the yield and tensile stresses of the oxidized nuclear fuel cladding tube decreased after 100 d of oxidation. The results of the scanning electron microscopy and transmission electron microscopy were represented with the in-situ Raman analysis result for the oxide characteristics generated on the cladding of spent nuclear fuel.

Zirconium alloy, Oxidation, Raman spectroscopy, Tensile test, Rupture test

The parent and daughter nuclides in a radioactive decay chain arrive at secular equilibrium once they have a large half-life difference. The characteristics of this equilibrium state can be used to estimate the production time of nuclear materials. In this study, a mathematical model and algorithm that can be applied to radio-chronometry using the radioactive equilibrium relationship were investigated, reviewed, and implemented. A Bateman equation that can analyze the decay of radioactive materials over time was used for the mathematical model. To obtain a differential-based solution of the Bateman equation, an algebraic numerical solution approach and two different matrix exponential functions (Moral and Levy) were implemented. The obtained result was compared with those of commonly used algorithms, such as the Chebyshev rational approximation method and WISE Uranium. The experimental analysis confirmed the similarity of the results. However, the Moral method led to an increasing calculation uncertainty once there was a branching decay, so this aspect must be improved. The time period corresponding to the production of nuclear materials or nuclear activity can be estimated using the proposed algorithm when uranium or its daughter nuclides are included in the target materials for nuclear forensics.

Nuclear forensics, Nuclear material, Radio chronometry, Decay chain, Isotopic signature

A compacted bentonite buffer is a major component of engineered barrier systems, which are designed for the disposal of high-level radioactive waste. In most countries, the target temperature required to maintain safe functioning is below 100°C. If the target temperature of the compacted bentonite buffer can be increased above 100°C, the disposal area can be dramatically reduced. To increase the target temperature of the buffer, it is necessary to investigate its properties at temperatures above 100°C. Although some studies have investigated thermal-hydraulic properties above 100°C, few have evaluated the water suction of compacted bentonite. This study addresses that knowledge gap by evaluating the water suction variation for compacted Korean bentonite in the 25–150°C range, with initial saturations of 0 and 0.22 under constant saturation conditions. We found that water suction decreased by 5–20% for a temperature increase of 100–150°C.

Compacted bentonite buffer, Elevated temperature condition, Water suction

Thermodynamically, TRUO_x, REO_x, and SrO_x can be chlorinated using ammonium chloride (NH₄Cl) as a chlorinating agent, whereas uranium oxides (U₃O₈ and UO₂) remain in the oxide form. In the preliminary experiments of this study, U₃O₈ and CeO₂ are reacted separately with NH₄Cl at 623 K in a sealed reactor. CeO₂ is highly reactive with NH₄Cl and becomes chlorinated into CeCl₃. The chlorination yield ranges from 96% to 100%. By contrast, U₃O₈ remains as UO₂ even after chlorination. We produced U/REO_x- and U/SrO_x-simulated fuels to understand the chlorination characteristics of the oxide compounds. Each simulated fuel is chlorinated with NH₄Cl, and the products are dissolved in LiCl-KCl salt to separate the oxide compounds from the chloride salt. The oxide compounds precipitate at the bottom. The precipitate and salt phases are sampled and analyzed via X-ray diffraction, scanning electron microscope-energy dispersive spectroscopy, and inductively coupled plasma-optical emission spectroscopy. The analysis results indicate that REO_x and SrO_x can be easily chlorinated from the simulated fuels; however, only a few of U oxide phases is chlorinated, particularly from the U/SrO_x-simulated fuels.

Uranium oxide, Transuranic oxide, Oxide spent nuclear fuel, Oxide chlorination, Ammonium chloride

This work provided a review of three techniques—(1) spectrochemical, (2) electrochemical, and (3) spectroelectrochemical–for molten salt medias. A spectroelectrochemical system was designed by utilizing this information. Here, we designed a spectroelectrochemical cell (SEC) and calibrated temperature controllers, and performed initial tests to explore the system’s capability limit. There were several issues and a redesign of the cell was accomplished. The modification of the design allowed us to assemble, align the system with the light sources, and successfully transferred the setup inside a controlled environment. A preliminary run was executed to obtain transmission and absorption background of NaCl-CaCl₂ salt at 600°C. It shows that the quartz cuvette has high transmittance effects across all wavelengths and there were lower transmittance effects at the lower wavelength in the molten salt media. Despite a successful initial run, the quartz vessel was mated to the inner cavity of the SEC body. Moreover, there was shearing in the patch cord which resulted in damage to the fiber optic cable, deterioration of the SEC, corrosion in the connection of the cell body, and fiber optic damage. The next generation of the SEC should attach a high temperature fiber optic patch cords without introducing internal mechanical stress to the patch cord body. In addition, MACOR should be used as the cell body materials to prevent corrosion of the surface and avoid the mating issue and a use of an adapter from a manufacturer that combines the free beam to a fiber optic cable should be incorporated in the future design.

Spectroelectrochemical method, Spectrochemical method, Electrochemical method, Molten salt

Several countries, including Korea, are considering the direct disposal of spent nuclear fuels. The radiological safety assessment results published after a geological repository closure indicate that the instant release is the main radiation source rather than the congruent release. Three Safety Case reports recently published were reviewed and the IRF values of seven long-lived radionuclides, including relevant experimental results, were compared. According to the literature review, the IRF values of both the CANDU and low burnup PWR spent fuel have been experimentally measured and used reasonably. In particular, the IRF values of volatile long-lived nuclides, such as ¹²⁹I and ¹³⁵Cs, were estimated from the FGR value. Because experimental leaching data regarding high burnup spent nuclear fuels are extremely scarce, a mathematical modelling approach proposed by Johnson and McGinnes was successfully applied to the domestic high burnup PWR spent nuclear fuel to derive the IRF values of iodine and cesium. The best estimate of the IRF was 5.5% at a discharge burnup of 55 GWd tHM⁻¹.

Disposal, Spent nuclear fuel, Instant release fraction, Fission gas release, High burnup

Spent filters with a high radiation dose rate of 2 mSv·hr⁻¹ or more are not easily managed. So far, the Korean policy for spent filter disposal is to store them temporarily at nuclear power plants until the waste filters can be easily managed. Nuclear power plant decommissioning in Korea is starting with Kori unit 1. Volume reduction of waste generated during decommissioning can reduce the cost and optimize the space usage at disposal site. Therefore, efficient volume reduction is a very important factor during the decommissioning process. A conceptual method, based on the experiences of developing 200 and 800 ton compactors at Orion EnC, has been developed considering worker exposure with the followings a crusher (upgrade of compaction efficiency), an automatic dose measuring system with a NaI(Tl) detector, a shield box, an inner drum to prepare for easy handling of drums and packaging, a 30 ton compactor, and an automatic robot system. This system achieves a volume reduction ratio of up to 85.7%; hence, the system can reduce the disposal cost and waste volume. It can be applied to other types of wastes that are not easily managed due to high dose rates and remote control operation necessity.

High dose, Spent filter, Handling system, Compaction equipment

Three government bodies, that is, the Ministry of Science and ICT (MSIT), Ministry of Trade, Industry, and Energy (MOTIE), and Nuclear Safety and Security (NSSC), jointly established the Institute for Korea Spent Nuclear Fuel (iKSNF) in December 2020 to secure the management technologies for spent nuclear fuel (SNF). The objective of iKSNF is to successfully conduct the long-term research and development program of the 「Development of Core Technologies to Ensure Safety of Spent Nuclear Fuel Storage and Disposal System」. Our program, known as the first multi-ministry program in the nuclear field of Korea, mainly focuses on developing core technologies required for the long-term management of SNF, including those for safe storage and deep geological disposal of SNF. The program comprises three subprograms and seven key projects covering the storage, disposal, and regulatory sectors of SNF management. Our program will last from 2021 through 2029, with a budget of approximately four billion USD sponsored by MSIT, MOTIE, and NSSC.

Spent nuclear fuel, Storage and transportation, Deep geological disposal

Molten salt consisting primarily of eutectic LiCl-KCl is currently being used in electrorefiners in the Fuel Conditioning Facility at Idaho National Laboratory. Options are currently being evaluated for storing this salt outside of the argon atmosphere hot cell. The hygroscopic nature of eutectic LiCl-KCl makes is susceptible to deliquescence in air followed by extreme corrosion of metallic cannisters. In this study, the effect of occluding the salt into a zeolite on water sorption/desorption was tested. Two zeolites were investigated: Na-Y and zeolite 4A. Na-Y was ineffective at occluding a high percentage of the salt at either 10 or 20wt% loading. Zeolite-4A was effective at occluding the salt with high efficiency at both loading levels. Weight gain in salt occluded zeolite-4A (SOZ) from water sorption at 20% relative humidity and 40℃ was 17wt% for 10% SOZ and 10wt% for 20% SOZ. In both cases, neither deliquescence nor corrosion occurred over a period of 31 days. After hydration, most of the water could be driven off by heating the hydrated salt occluded zeolite to 530℃. However, some HCl forms during dehydration due to salt hydrolysis. Over a wide range of temperatures (320–700℃) and ramp rates (5, 10, and 20℃ min⁻¹), HCl formation was no more than 0.6% of the Cl⁻ in the original salt.

Zeolite, Electrorefiner salt, Ceramic waste form, Hydrolysis

The buffer is a critical barrier component in an engineered barrier system, and its purpose is to prevent potential radionuclides from leaking out from a damaged canister by filling the void in the repository. No experimental parameters exist that can describe the buffer expansion phenomenon when Kyeongju bentonite, which is a buffer candidate material available in Korea, is exposed to groundwater. As conventional experiments to determine these parameters are time consuming and complicated, simple swelling pressure tests, numerical modeling, and machine learning are used in this study to obtain the parameters required to establish a numerical model that can simulate swelling. Swelling tests conducted using Kyeongju bentonite are emulated using the COMSOL Multiphysics numerical analysis tool. Relationships between the swelling phenomenon and mechanical parameters are determined via an artificial neural network. Subsequently, by inputting the swelling tests results into the network, the values for the mechanical parameters of Kyeongju bentonite are obtained. Sensitivity analysis is performed to identify the influential parameters. Results of the numerical analysis based on the identified mechanical parameters are consistent with the experimental values.

Bentonite, Sensitivity analysis, Artificial neural network, Parameter identification

The fundamental characteristics of groundwater colloids, such as composition, concentration, size, and stability, were analyzed using granitic groundwater samples taken from the KAERI Underground Research Tunnel (KURT) site by such analytical methods as inductively coupled plasma-mass spectrometry, field emission-transmission electron microscopy, a liquid chromatography-organic carbon detector, and dynamic light scattering technique. The results show that the KURT groundwater colloids are mainly composed of clay minerals, calcite, metal (Fe) oxide, and organic matter. The size and concentration of the groundwater colloids were 10–250 nm and 33–64 μg·L⁻¹, respectively. These values are similar to those from other studies performed in granitic groundwater. The groundwater colloids were found to be moderately stable under the groundwater conditions of the KURT site. Consequently, the groundwater colloids in the fractured granite system of the KURT site can form stable radiocolloids and increase the mobility of radionuclides if they associate with radionuclides released from a radioactive waste repository. The results provide basic data for evaluating the effects of groundwater colloids on radionuclide migration in fractured granite rock, which is necessary for the safety assessment of a high-level radioactive waste repository.

Granite, Groundwater colloid, Inorganic colloid, Organic colloid, Stability

The acoustic emission (AE) is proposed as a feasible method for the real-time monitoring of the structural damage evolution in concrete materials that are typically used in the storage of nuclear wastes. However, the characteristics of AE signals emitted from concrete structures subjected to various environmental conditions are poorly identified. Therefore, this study examines the AE characteristics of the concrete structures during uniaxial compression, where the storage temperature and immersion conditions of the concrete specimens varied from 15℃ to 75℃ and from completely dry to water-immersion, respectively. Compared with the dry specimens, the water-immersed specimens exhibited significantly reduced uniaxial compressive strengths by approximately 26%, total AE energy by approximately 90%, and max RA value by approximately 70%. As the treatment temperature increased, the strength and AE parameters, such as AE count, AE energy, and RA value, of the dry specimens increased; however, the temperature effect was only minimal for the immersed specimens. This study suggests that the AE technique can capture the mechanical damage evolution of concrete materials, but their AE characteristics can vary with respect to the storage conditions.

Nuclear waste storage, Acoustic emission, Concrete, Storage condition, Damage

60+ Years of nuclear power generation has led to a significant legacy of radioactively contaminated land at a number of nuclear licenced “mega sites” around the world. The safe management and remediation of these sites is key to ensuring there environmental stewardship in the long term. Bioremediation utilizes a variety of microbially mediated processes such as, enzymatically driven metal reduction or biominerialisation, to sequester radioactive contaminants from the subsurface limiting their migration through the geosphere. Additionally, some of these process can provide environmentally stable sinks for radioactive contaminants, through formation of highly insoluble mineral phases such as calcium phosphates and carbonates, which can incorporate a range of radionuclides into their structure. Bioremediation options have been considered and deployed in preference to conventional remediation techniques at a number of nuclear “mega” sites. Here, we review the applications of bioremediation technologies at three key nuclear licenced sites; Rifle and Hanford, USA and Sellafield, UK, in the remediation of radioactively contaminated land.

Bioremediation, Remediation, Geomicrobiology, Nuclear Sites, Environmental

The transport of radionuclides at oceanic scales can be assessed using a Lagrangian model. In this review an application of such a model to the Atlantic, Indian and Pacific oceans is described. The transport model, which is fed with water currents provided by global ocean circulation models, includes advection by three–dimensional currents, turbulent mixing, radioactive decay and adsorption/release of radionuclides between water and bed sediments. Adsorption/release processes are described by means of a dynamic model based upon kinetic transfer coefficients. A stochastic method is used to solve turbulent mixing, decay and water/sediment interactions. The main results of these oceanic radionuclide transport studies are summarized in this paper. Particularly, the potential leakage of ¹³⁷Cs from dumped nuclear wastes in the north Atlantic region was studied. Furthermore, hypothetical accidents, similar in magnitude to the Fukushima accident, were simulated for nuclear power plants located around the Indian Ocean coastlines. Finally, the transport of radionuclides resulting from the release of stored water, which was used to cool reactors after the Fukushima accident, was analyzed in the Pacific Ocean.

Dispersion, Ocean circulation, Lagrangian model, Radioactive release

Major accidents at nuclear power plants generate huge amounts of radioactive waste in a short period of time over a wide area outside the plant boundary. Therefore, extraordinary efforts are required for safe management of the waste. A well-established remediation plan including radioactive waste management that is prepared in advance will minimize the impact on the public and environment. In Korea, however, only limited plans exist to systematically manage this type of off-site radioactive waste generating event. In this study, we developed basic strategies for off-site radioactive waste management based on recommendations from the IAEA (International Atomic Energy Agency) and NCRP (National Council on Radiation Protection and Measurements), experiences from the Fukushima Daiichi accident in Japan, and a review of the national radioactive waste management system in Korea. These strategies included the assignment of roles and responsibilities, development of management methodologies, securement of storage capacities, preparation for the use of existing infrastructure, assurance of information transparency, and establishment of cooperative measures with international organizations.

Major accident, Fukushima Daiichi accident, Nuclear power plant, Radioactive waste, Management plan, Countermeasures

Nuclear power plant decommissioning generates significant concrete waste, which is slightly contaminated, and expected to be classified as clearance concrete waste. Clearance concrete waste is generally crushed into rubble at the site or a satellite treatment facility for practical disposal purposes. During the process, workers are exposed to radiation from the nuclides in concrete waste. The treatment processes consist of concrete cutting/crushing, transportation, and loading/unloading. Workers’radiation exposure during the process was systematically studied. A shielding package comprising a cylindrical and hexahedron structure was considered to reduce workers’ radiation exposure, and improved the treatment process’s efficiency. The shielding package’s effect on workers’ radiation exposure during the cutting and crushing process was also studied. The calculated annual radiation exposure of concrete treatment workers was below 1 mSv, which is the annual radiation exposure limit for members of the public. It was also found that workers involved in cutting and crushing were exposed the most.

Radioactive waste, Concrete, Clearance, Radiation Exposure, Shielding

The safety of a KTC-360 transport cask, a large-capacity pressurized heavy-water reactor transport cask that transports CANDU spent nuclear fuel discharged from the reactor after burning in a pressurized heavy-water reactor, must be demonstrated under the normal transport and accident conditions specified under transport cask regulations. To confirm the thermal integrity of this cask under normal transport and accident conditions, high-temperature and fire tests were performed using a one-third slice model of an actual KTC-360 cask. The results revealed that the surface temperature of the cask was 62°C, indicating that such casks must be transported separately. The highest temperature of the CANDU spent nuclear fuel was predicted to be lower than the melting temperature of Zircaloy-4, which was the sheath material used. Therefore, if normal operating conditions are applied, the thermal integrity of a KTC-360 cask can be maintained under normal transport conditions. The fire test revealed that the maximum temperatures of the structural materials, stainless steel, and carbon steel were 446°C lower than the permitted maximum temperatures, proving the thermal integrity of the cask under fire accident conditions.

Pressurized heavy water reactor, High-temperature test, Fire test, Thermal integrity, Transport cask

Operating and decommissioning nuclear power plants generates radioactive waste. This radioactive waste can be categorized into several different levels, for example, low, intermediate, and high, according to the regulations. Currently, low and intermediate-level waste are stored in conventional 200-liter drums to be disposed. However, in Korea, the disposal of intermediate-level radioactive waste is virtually impossible as there are no available facilities. Furthermore, large-sized intermediate-level radioactive waste, such as reactor internals from decommissioning, need to be segmented into smaller sizes so they can be adequately stored in the conventional drums. This segmentation process requires additional costs and also produces secondary waste. Therefore, this paper suggests repurposing the no-longer-used spent nuclear fuel casks. The casks are larger in size than the conventional drums, thus requiring less segmentation of waste. Furthermore, the safety requirements of the spent nuclear fuel casks are severer than those of the drums. Hence, repurposed spent nuclear fuel casks could better address potential risks such as dropping, submerging, or a fire. In addition, the spent nuclear fuel casks need to be disposed in compliance with the regulations for low level radioactive waste. This cost may be avoided by repurposing the casks.

Radioactive waste storage, Decommissioning, Spent nuclear fuel cask, Repurposing

Carbonates are inorganic ligands that are abundant in natural groundwater. They strongly influence radionuclide mobility by forming strong complexes, thereby increasing solubility and reducing soil absorption rates. We characterized the spectroscopic properties of Am(III)-carbonate species using UV-Vis absorption and time-resolved laser-induced fluorescence spectroscopy. The deconvoluted absorption spectra of aqueous Am(CO₃)₂⁻ and Am(CO₃)₃ ³⁻ species were identified at redshifted positions with lower molar absorption coefficients compared to the absorption spectrum of aqua Am³⁺. The luminescence spectrum of Am(CO₃)₃ ³⁻ was red-shifted from 688 nm for Am³⁺ to 695 nm with enhanced intensity and an extended lifetime. Colloidal Am(III)-carbonate compounds exhibited absorption at approximately 506 nm but had non-luminescent properties. Slow formation of colloidal particles was monitored based on the absorption spectral changes over the sample aging time. The experimental results showed that the solubility of Am(III) in carbonate solutions was higher than the predicted values from the thermodynamic constants in OECD-NEA reviews. These results emphasize the importance of kinetic parameters as well as thermodynamic constants to predict radionuclide migration. The identified spectroscopic properties of Am(III)-carbonate species enable monitoring time-dependent species evolution in addition to determining the thermodynamics of Am(III) in carbonate systems.

Americium, Carbonate, Radionuclide migration, Spectroscopy, Thermodynamics, Kinetics

The electrochemical behavior was investigated during the electrolysis of nickel oxide in LiCl-Li₂O salt mixture at 650℃by changing several components. The focus of this work is to improve anode design and shroud design to increase current densities. The tested components were ceramic anode shroud porosity, porosity size, anode geometry, anode material, and metallic porous anode shroud. The goal of these experiments was to optimize and improve the reduction process. The highest contributors to higher current densities were anode shroud porosity and anode geometry.

Oxide reduction, Electrolysis, Molten Salt, Process optimization, Nickel oxide

The solubility and species distribution of radionuclides in groundwater are essential data for the safety assessment of deep underground spent nuclear fuel (SNF) disposal systems. Americium is a major radionuclide responsible for the long-term radiotoxicity of SNF. In this study, the solubility of americium compounds was evaluated in synthetic groundwater (Syn-DB3), simulating groundwater from the DB3 site of the KAERI Underground Research Tunnel. Geochemical modeling was performed using the ThermoChimie_11a thermochemical database. Concentration of dissolved Am(III) in Syn-DB3 in the pH range of 6.4–10.5 was experimentally measured under over-saturation conditions by liquid scintillation counting over 70 d. The absorption spectra recorded for the same period suggest that Am(III) colloidal particles formed initially followed by rapid precipitation within 2 d. In the pH range of 7.5–10.5, the concentration of dissolved Am(III) converged to approximately 2×10⁻⁷ M over 70 d, which is comparable to that of the amorphous AmCO₃OH(am) according to the modeling results. As the samples were aged for 70 d, a slow equilibrium process occurred between the solid and solution phases. There was no indication of transformation of the amorphous phase into the crystalline phase during the observation period.

Americium, Groundwater, Radionuclide migration, Solubility, Colloidal particle, Thermochemical data

Given the domestic situation, all nuclear power plants are located at the seaside, where interim storage sites are also likely to be located and maritime transportation is considered inevitable. Currently, Korea does not have an independently developed maritime transportation risk assessment code, and no research has been conducted to evaluate the release rate of radioactive waste from a submerged transportation cask in the sea. Therefore, secure technology is necessary to assess the impact of immersion accidents and establish a regulatory framework to assess, mitigate, and prevent maritime transportation accidents causing serious radiological consequences. The flow rate through a gap in a containment boundary should be calculated to determine the accurate release rate of radionuclides. The fluid flow through the micro-scale gap can be evaluated by combining the flow inside and outside the transportation cask. In this study, detailed computational fluid dynamic and simplified models are constructed to evaluate the internal flow in a transportation cask and to capture the flow and heat transfer around the transportation cask in the sea, respectively. In the future, fluid flow through the gap will be evaluated by coupling the models developed in this study.

Spent nuclear fuel, Barrier effect model, Maritime transportation, Risk assessment, Release rate

Numerical modeling and scenario composition are needed to characterize the geological environment of the disposal site and analyze the long-term evolution of natural barriers. In this study, processes and features of the hydro-mechanical behavior of natural barriers were categorized and represented using the interrelation matrix proposed by SKB and Posiva. A hydro-mechanical coupled model was evaluated for analyzing stress field changes and fracture zone re-activation. The processes corresponding to long-term evolution and the hydro-mechanical mechanisms that may accompany critical processes were identified. Consequently, practical numerical methods could be considered for these geological engineering issues. A case study using a numerical method for the stability analysis of an underground disposal system was performed. Critical stress distribution regime problems were analyzed numerically by considering the strata’s movement. Another case focused on the equivalent continuum domain composition under the upscaling process in fractured rocks. Numerical methods and case studies were reviewed, confirming that an appropriate and optimized modeling technique is essential for studying the stress state and geological history of the Korean Peninsula. Considering the environments of potential disposal sites in Korea, selecting the optimal application method that effectively simulates fractured rocks should be prioritized.

Geological disposal, Natural barriers, Long-term evolution, Numerical modeling, Hydro-mechanical coupling

Colloid Formation and Migration (CFM) project is being carried out within the Grimsel Test Site (GTS) Phase Ⅵ. Since 2008, the Korea Atomic Energy Research Institute (KAERI) has joined CFM to investigate the behavior of colloid-facilitated radionuclide transport in a generic Underground Research Laboratory (URL). The CFM project includes a long-term in-situ test (LIT) and an in-rock bentonite erosion test (i-BET) to assess the in-situ colloid-facilitated radionuclide transport through the bentonite erosion in the natural flow field. In the LIT experiment, radionuclide-containing compacted bentonite was equipped with a triple-packer system and then positioned at the borehole in the shear zone. It was observed that colloid transport was limited owing to the low swelling pressure and low hydraulic conductivity. Therefore, a postmortem analysis is being conducted to estimate the partial migration and diffusion of radionuclides. The i-BET experiment, that focuses more on bentonite erosion, was newly designed to assess colloid formation in another flow field. The i-BET experiment started with the placement of compacted bentonite rings in the double-packer system, and the hydraulic parameters and bentonite erosion have been monitored since December 2018.

Colloid, Grimsel test site, Colloid formation and migration, Long-term in-situ test, in-rock bentonite erosion test

Technology for high-level-waste disposal employing a multibarrier concept using engineered and natural barrier in stable bedrock at 300–1,000 m depth is being commercialized as a safe, long-term isolation method for high-level waste, including spent nuclear fuel. Managing heat generated from waste is important for improving disposal efficiency; thus, research on efficient heat management is required. In this study, thermal management methods to maximize disposal efficiency in terms of the disposal area required were developed. They efficiently use the land in an environment, such as Korea, where the land area is small and the amount of waste is large. The thermal effects of engineered barriers and natural barriers in a high-level waste disposal repository were analyzed. The research status of thermal management for the main bedrocks of the repository, such as crystalline, clay, salt, and other rocks, were reviewed. Based on a characteristics analysis of various heat management approaches, the spent nuclear fuel cooling time, buffer bentonite thermal conductivity, and disposal container size were chosen as efficient heat management methods applicable in Korea. For each method, thermal analyses of the disposal repository were performed. Based on the results, the disposal efficiency was evaluated preliminarily. Necessary future research is suggested.

Spent nuclear fuels, High-level waste, Geological disposal, Decay heat, Thermal management, Bentonite block, Disposal container, Engineered barrier, Natural barrier

The decommissioning of nuclear facilities produces various types of radiologically contaminated waste. In addition, dismantlement activities, including cutting, packing, and clean-up at the facility site, result in secondary radioactive waste such as filters, resin, plastic, and clothing. Determining of the radionuclide content of this waste is an important step for the determination of a suitable management strategy including classification and disposal. In this work, we radiochemically characterized the radionuclide activities of filters used during the decommissioning of Korea Research Reactors (KRRs) 1 and 2. The results indicate that the filter samples contained mainly ³H (500–3,600 Bq·g⁻¹), ¹⁴C (7.5–29 Bq·g⁻¹), ⁵⁵Fe (1.1–7.1 Bq·g⁻¹), ⁵⁹Ni (0.60–1.0 Bq·g⁻¹), ⁶⁰Co (0.74–70 Bq·g⁻¹), ⁶³Ni (0.60–94 Bq·g⁻¹), ⁹⁰Sr (0.25–5.0 Bq·g⁻¹), ¹³⁷Cs (0.64–8.7 Bq·g⁻¹), and ¹⁵²Eu (0.19–2.9) Bq·g⁻¹. In addition, the gross alpha radioactivity of the samples was measured to be between 0.32–1.1 Bq·g⁻¹. The radionuclide concentrations were below the concentration limit stated in the low- and intermediatelevel waste acceptance criteria of the Nuclear Safety and Security Commission, and used for the disposal of the KRRs waste drums to a repository site.

Characterization of radioactive waste, Inventory of radionuclides, Radiochemistry, Difficult to measure

The nuclear criticality analyses considering burnup credit were performed for a spent nuclear fuel (SNF) disposal cell consisting of bentonite buffer and two different types of SNF disposal canister: the KBS-3 canister and small standardized transportation, aging and disposal (STAD) canister. Firstly, the KBS-3 & STAD canister containing four SNFs of the initial enrichment of 4.0wt% ²³⁵U and discharge burnup of 45,000 MWD/MTU were modelled. The k_eff values for the cooling times of 40, 50, and 60 years of SNFs were calculated to be 0.79108, 0.78803, and 0.78484 & 0.76149, 0.75683, and 0.75444, respectively. Secondly, the KBS-3 & STAD canister with four SNFs of 4.5wt% and 55,000 MWD/MTU were modelled. The k_eff values for the cooling times of 40, 50, and 60 years were 0.78067, 0.77581, and 0.77335 & 0.75024, 0.74647, and 0.74420, respectively. Therefore, all cases met the performance criterion with respect to the k_eff value, 0.95. The STAD canister had the lower k_eff values than KBS-3. The neutron absorber plates in the STAD canister significantly affected the reduction in k_eff values although the distance among the SNFs in the STAD canister was considerably shorter than that in the KBS-3 canister.

Nuclear criticality analysis, Burnup credit, Disposal cell, KBS-3 disposal canister, STAD disposal canister, Spent nuclear fuel

Decommissioning of nuclear power plants generates a large amount of radioactive waste in a short period. Moreover, Radioactive waste has various forms including a large volumes of metal, concrete, and solid waste. The disposal of decommissioning waste using 200 L drums is inefficient in terms of economics, work efficiency, and radiation safety. Therefore, The Korea Radioactive Waste Agency is developing large containers for the packaging, transportation, and disposal of decommissioning waste. Assessing disposability considering the characteristics of the radioactive waste and facility, convenience of operation, and safety of workers is necessary. In this study, the exposure dose rate of workers during the disposal of new containers was evaluated using Monte Carlo N-Particle Transport code. Six normal and four abnormal scenarios were derived for the assessment of the dose rate in a near surface disposal facility operation. The results showed that the calculated dose rates in all normal scenarios were lower than the direct exposure dose limitation of workers in the safety analysis report. In abnormal scenarios, the work hours with dose rates below 20 mSv·y⁻¹ were calculated. The results of this study will be useful in establishing the optimal radiation work conditions.

Decommissioning radioactive waste, Radiation worker, Disposal container, Exposure dose, Working hours, Near surface disposal facility

The International Atomic Energy Agency (IAEA) entails independent decision-making for the safety supervision of civil nuclear facilities. To evaluate and review the safety of nuclear facilities, the national regulatory body usually consults independent institutions or external committees. Technical Support Organizations (TSOs) include national laboratories, research institutions, and consulting organizations. Support from professional organizations in other countries may also be required occasionally. Most of the world’s major nuclear power countries adopt an independent nuclear safety supervision model. Accordingly, China has continuously improved upon the construction of such a system by establishing the National Nuclear Safety Administration (NNSA) as the decision-making department for nuclear and radiation safety supervision, six regional safety supervision stations, the Nuclear and Radiation Safety Center (NSC), a nuclear safety expert committee, and the National Nuclear and Radiation Safety Supervision Technology R&D Base, which serves as the test, verification, and R&D platform for providing consultation and technical support. An R&D system, however, remains to be formed. Future endeavors must focus on improving the technical support capacity of these systems. As an enhancement from institutional independence to capability independence is necessary for ensuring the independence of China’s nuclear safety regulatory institution, its regulatory capacity must be improved in the future.

Nuclear facilities, Supervision, Technical support, Test bench, Safety analysis

Because of the massive development of nuclear power plants in China in recent years, China is facing the challenge of radioactive waste disposal. China has established complete regulatory requirements for radioactive waste disposal, but it also has encountered problems and challenges in low-level radioactive waste disposal in terms of management, selection of disposal facility sites, and implementation of a site selection plan. Three low-level radioactive waste disposal facilities that have been operated in China are described, and their activity limits, locations, and capacities are also outlined. The connotations of “regional” and “centralized” disposal policies are discussed in light of the characteristics of the radioactive waste. The characteristics and advantages of the regional and centralized disposal policies are compared. It is concluded that the regional disposal policy adopted in 1992 can no longer meet the current disposal needs, and China should adopt a combination of the two disposal policies to solve the problem of radioactive waste disposal.

Disposal, Nuclear power, Waste

When decommissioning nuclear power plant (NPP), the first task performed is cost estimation. This is an important task in terms of securing adequate decommissioning funds and managing the schedule. Therefore, many countries and institutions are conducting continuous research and also developing and using many programs for cost estimation. However, the cost estimated for decommissioning an NPP typically differs from the actual cost incurred in its decommissioning. This is caused by insufficient experience in decommissioning NPPs or lack of decommissioning cost data. This uncertainty in cost estimation can be in general compensated for by applying a contingency. However, reflecting an appropriate standard for the contingency is also difficult. Therefore, in this study, data analysis was conducted based on the contingency guideline suggested by each institution and the actual cost of decommissioning the NPP. Subsequently, TLG Service, Inc.’s process, which recently suggested specific decommissioning costs, was matched with ISDC (International Structure for Decommissioning Costing)’s work breakdown structure (WBS). Based on the matching result, the guideline for applying the contingency for ISDC’s WBS Level 1 were presented. This study will be helpful in cost estimation by applying appropriate contingency guidelines in countries or institutions that have no experience in decommissioning NPPs.

Cost estimation, Contingency, Decommissioning, ISDC, WBS

The pH dependence of sorption distribution coefficient (K_d) of Np(IV) on MX-80 in Ca-Na-Cl type solution with the ionic strength of 0.3 M, which was similar to one of the reference groundwaters in crystalline rock, was experimentally investigated under the reducing conditions. The overall trend of K_d on MX-80 was independent of pH at 5 ≤ pH ≤ 10 but increased as pH increased at pH ≤ 5. The 2-site protolysis non-electrostatic surface complexation and cation exchange model was applied to the experimentally measured pH dependence of K_d and the optimized surface complexation constants of Np(IV) sorption on MX-80 were estimated. The values of surface complexation constants in this work agreed relatively well with those in the Na-Ca-Cl solution previously evaluated, suggesting that compared to Na⁺, the competition of Ca²⁺ with Np(IV) for surface complexation on MX-80 was not much strong in Ca-Na-Cl solution. The sorption model well predicted the pH dependence of K_d values but slightly overestimated the sorption at the low pH region.

Np(IV),MX-80,Sorption,pH dependence,Sorption distribution coefficient,2 SPNE SC/CE model

As an example of research activities in decontamination for decommissioning, new data are presented on the options for corrosion layer dissolution during the decommissioning decontamination, or persulfate regeneration for decontamination solutions re-use. For the management of spent decontamination solutions, new method based on solvent extraction of radionuclides into ionic liquid followed by electrodeposition of the radionuclides has been developed. Fields of applications of composite inorganic-organic absorbers or solid extractants with polyacrylonitrile (PAN) binding matrix for the treatment of liquid radioactive waste are reviewed; a method for americium separation from the boric acid containing NPP evaporator concentrates based on the TODGA-PAN material is discussed in more detail. Performance of a model of radionuclide transport, developed and implemented within the GoldSim programming environment, for the safety studies of the LLW/ILW repository is demonstrated on the specific case of the Richard repository (Czech Republic). Continuation and even broadening of these activities are expected in connection with the approaching end of the lifespan of the first blocks of the Czech NPPs.

Liquid radioactive waste,Decontamination,Ionic liquids,Composite absorbers,Solid extractants,Safety assessment

The article summarises the status and competence of UJV Rez, a. s. (up to 2012, the Nuclear Research Institute Rez, Czech Republic) in the field of radioactive waste (RAW) management as a company managing of 95% of institutional radioactive wastes in Czech Republic. UJV Rez a. s. has been one of the Czech Republic’s key research and engineering institutions in the field of nuclear energy production since 1955. The company processes and conditions prior to storage 95% of so-called institutional RAW and is the principal partner of the state with respect to the research support of the Czech deep geological repository development project. UJV Rez a. s. boasts its own accredited radiochemical analytical test laboratory, unique of its kind in the Czech Republic. Of equal importance is UJV Rez’s active participation in a range of international organisations and associations and its involvement in wide range of international projects, and so as European projects. One of them is EU funded project PREDIS: Pre-disposal management of radioactive wastes, that has started at September 2020, focused on the field of low level radioactive waste (LLW) and intermediate level radioactive waste (ILW) pre-disposal.

Bitumen,Concrete,Conditioning,Pre-disposal,Radioactive waste,Radioactive waste management

The behaviors of various desorption agents were investigated during the desorption of cesium (Cs) from samples of clay minerals and actual soil. Results showed that polymeric cation exchange agents (polyethyleneimine (PEI)) efficiently desorbed Cs from expandable montmorillonite, whereas acidic desorption solutions containing HCl or PEI removed considerable Cs from hydrobiotite. However, most desorption agents could desorb only 54% of Cs from illite because of Cs’s specific adsorption to selective adsorption sites. Cs desorption from an actual soil sample containing Cs-selective clay mineral illite (< 200 μm) and extracted from near South Korea’s Kori Nuclear Power Plant was also investigated. Considerable adsorbed ¹³⁷Cs was expected to be located at Cs-selective sites when the ¹³⁷Cs loading was much lower than the sample’s cation exchange capacity. At this low ¹³⁷Cs loading, the total Cs amount desorbed by repeated washing varied by desorption agent in the order HCl > PEI > NH₄⁺, and the highest Cs desorption amount achieved using HCl was 83%. Unlike other desorption agents with only cation exchange capabilities, HCl can attack minerals and induce dissolution of metallic elements. HCl’s ability to both alter minerals and induce H⁺/Cs⁺ ion exchange is expected to promote Cs desorption from actual soil samples.

Cesium,Soil,Clay,Desorption,Washing

Chemical decontamination of primary systems in a nuclear power plant (NPP) prior to commencing the main decommissioning activities is required to reduce radiation exposure during its process. The entire process is repeated until the desired decontamination factor is obtained. To achieve improved decontamination factors over a shorter time with fewer cycles, the appropriate flow characteristics are required. In addition, to prepare an operating procedure that is adaptable to various conditions and situations, the transient analysis results would be required for operator action and system impact assessment. In this study, the flow characteristics in the steady-state and transient conditions for the chemical decontamination operations of the Kori-1 NPP were analyzed and compared via the MARS-KS code simulation. Loss of residual heat removal (RHR) and steam generator tube rupture (SGTR) simulations were conducted for the postulated abnormal events. Loss of RHR results showed the reactor coolant system (RCS) temperature increase, which can damage the reactor coolant pump (RCP)s by its cavitation. The SGTR results indicated a void formation in the RCS interior by the decrease in pressurizer (PZR) pressure, which can cause surface exposure and tripping of the RCPs unless proper actions are taken before the required pressure limit is achieved.

Kori-1 Nuclear Power Plant,Decommissioning,Chemical decontamination,Flow characteristics,Transient analysis,MARS-KS

A paradigm shift in the government’s energy policy was reflected in its declaration of early closure of old nuclear plants as well as cancellation of plans for the construction of new plants. To this end, unit 1 of Kori Nuclear Power Plant was permanently shut down and is set for decommission. Based on these changes, the off-site transport of spent fuels from nuclear power plants has become a critical issue. The purpose of this study is to develop an optimized method for transportation of spent fuels from Kori Nuclear Power Plant’s units 1, 2, 3, and 4 to an assumed interim storage facility by simulating the scenarios using the Flexsim software, which is widely used in logistics and manufacturing applications. The results of the simulation suggest that the optimized transport methods may contribute to the development of delivery schedule of spent fuels in the near future. Furthermore, these methods can be applied to decommissioning plan of nuclear power plants.

Kori NPP,Nuclear power plant decommissioning,Transportation of spent fuels,Logistics,Discrete event,Discrete simulation

Evaluating the quantitative damage to rocks through acoustic emission (AE) has become a research focus. Most studies mainly used one or two AE parameters to evaluate the degree of damage, but several AE parameters have been rarely used. In this study, several data-driven models were employed to reflect the combined features of AE parameters. Through uniaxial compression tests, we obtained mechanical and AE-signal data for five granite specimens. The maximum amplitude, hits, counts, rise time, absolute energy, and initiation frequency expressed as the cumulative value were selected as input parameters. The result showed that gradient boosting (GB) was the best model among the support vector regression methods. When GB was applied to the testing data, the root-mean-square error and R between the predicted and actual values were 0.96 and 0.077, respectively. A parameter analysis was performed to capture the parameter significance. The result showed that cumulative absolute energy was the main parameter for damage prediction. Thus, AE has practical applicability in predicting rock damage without conducting mechanical tests. Based on the results, this study will be useful for monitoring the near-field rock mass of nuclear waste repository.

Rock damage,Acoustic emission,Nuclear waste repository,Gradient boosting,Support vector regression

Nuclear power plants (NPPs) produce radioactive waste and decommissioning this waste entails additional cost; determining these costs for various types and specifications of radioactive waste can be challenging. The purpose of this study is to identify major determinants of the decommissioning cost and their impact on NPPs. To this end, data from defunct NPPs were gathered and 2SLS (Two Stage Least Squares) regression models were developed to investigate the major contributors depending on the reactor types, viz. PWR (Pressurized Water Reactors) and BWR (Boiling Water Reactors). Additionally, cost estimations and the Monte Carlo simulation were performed as part of performance validation. Our study established that the decommissioning costs are primarily influenced by the level of radioactivity in the decommissioned waste, which can be realized from operational factors like operation period, overall efficiency, and plant capacity, as well as from duration of decommissioning and labour cost. While our study provides an improved statistical approach to recognize these factors, we acknowledge that our models have limitations in forecasting accurately which we envisage to bolster in future studies by identifying more substantive factors.

Nuclear power plants,Decommissioning cost,Radioactive waste,2SLS regression

A safety assessment of radioactive waste repositories is a mandatory requirement process because there are possible radiological hazards owing to radionuclide migration from radioactive waste to the biosphere. For a reliable safety assessment, it is important to establish a parameter database that reflects the site-specific characteristics of the disposal facility and repository site. From this perspective, solubility, a major geochemical parameter, has been chosen as an important parameter for modeling the migration behavior of radionuclides. The solubilities were derived for Am, Ni, Tc, and U, which were major radionuclides in this study, and on-site groundwater data reflecting the operational conditions of the Gyeongju low and intermediate level radioactive waste (LILW) repository were applied to reflect the site-specific characteristics. The radiation dose was derived by applying the solubility and radionuclide inventory data to the RESRAD-OFFSITE code, and sensitivity analysis of the dose according to the solubility variation was performed. As a result, owing to the low amount of radionuclide inventory, the dose variation was insignificant. The derived solubility can be used as the main input data for the safety assessment of the Gyeongju LILW repository in the future.

Radioactive waste disposal facility,Safety assessment,Geochemical parameter,Solubility

Radiological impact analyses were carried out for a near-surface radioactive waste repository at Gyeongju in South Korea. The RESRAD-ONSITE code was applied for the estimation of maximum exposure doses by considering various exposure pathways based on a land area of 2,500 m² with a 0.15 m thick contamination zone. Typical influencing input parameters such as shield depth, shield materials’ density, and shield erosion rate were examined for a sensitivity analysis. Then both residential farmer and industrial worker scenarios were used for the estimation of maximum exposure doses depending on exposure duration. The radiation dose evaluation results showed that ⁶⁰Co, ¹³⁷Cs, and ⁶³Ni were major contributors to the total exposure dose compared with other radionuclides. Furthermore, the total exposure dose from ingestion (plant, meat, and milk) of the contaminated plants was more significant than those assessed for inhalation, with maximum values of 5.5×10⁻⁴ mSv‧yr⁻¹ for the plant ingestion. Thus the results of this study can be applied for determining near-surface radioactive waste repository conditions and providing quantitative analysis methods using RESRAD-ONSITE code for the safety assessment of disposing radioactive materials including decommissioning wastes to protect human health and the environment.

Radiation impact,Radioactive waste repository,Radionuclides,RESRAD,Safety assessment

This study evaluates the radioactivity of concrete waste that occurs due to large amounts of decommissioned nuclear wastes and then determines the surface dose rate when the waste is packaged in a disposal container. The radiation assessment was conducted under the presumption that impurities included in the bio-shielded concrete contain the highest amount of radioactivity among all the concrete wastes. Neutron flux was applied using the simplified model approach in a sample containing the most Co and Eu impurities, and a maximum of 9.8×10⁴ Bq·g^{−1 60}Co and 2.63×10⁵ Bq·g^{−1 152}Eu was determined. Subsequently, the surface dose rate of the container was measured assuming that the bio-shield concrete waste would be packaged in a newly developed disposal container. Results showed that most of the concrete wastes with a depth of 20 cm or higher from the concrete surface was found to have less than 1.8 mSv·hr⁻¹ in the surface dose of the new-type disposal container. Hence, when bio-shielded concrete wastes, having the highest radioactivity, is disposed in the new disposal container, it satisfies the limit of the surface dose rate (i.e., 2 mSv·hr⁻¹) as per global standards.

LILW,Radioactive evaluation,Decommissioned concrete waste,Disposal container

Unit 1 of the Kori Nuclear Power Plant (NPP) and Unit 1 of the Wolsong NPP are being prepared for decommissioning; their decommissioning is expected to generate large amounts of intermediate-level, low-level, and very low level Waste. Mixed waste containing both radioactive and hazardous substances is expected to be produced. Nevertheless, laws and regulations, such as the Korean Nuclear Safety Act and Waste Management Act, do not define clear regulatory guidelines for mixed waste. However, the United States has strictly enforced regulations on mixed waste, focusing on the human health and environmental effects of its hazardous components. The U.S. Nuclear Regulatory Commission and the U.S. Department of Energy regulate the radioactive components of mixed waste under the Atomic Energy Act. The U.S. Environmental Protection Agency regulates the hazardous waste component of mixed waste under the Resource Conservation and Recovery Act. In this study, the laws, regulations, and authorities pertaining to mixed waste in the United States are reviewed. Through comparison and analysis with waste management laws and regulations in Korea, a treatment direction for mixed waste is suggested. Such a treatment for mixed waste will increase the efficiency of managing mixed waste when decommissioning NPPs in the near future.

Kori-1,Nuclear power plant decommissioning,Waste management,Mixed waste,Laws and regulations

In this study, the electrochemical behavior of Sm on the binary liquid Al-Ga cathode in the LiCl-KCl molten salt system is investigated. First, the co-reduction process of Sm(III)-Al(III), Sm(III)-Ga(III), and Sm(III)-Ga(III)-Al(III) on the W electrode (inert) were studied using cyclic voltammetry (CV), square-wave voltammetry (SWV) and open circuit potential (OCP) methods, respectively. It was identified that Sm(III) can be co-reduced with Al(III) or Ga(III) to form Al_zSm_y or Ga_xSm_y intermetallic compounds. Subsequently, the under-potential deposition of Sm(III) at the Al, Ga, and Al-Ga active cathode was performed to confirm the formation of Sm-based intermetallic compounds. The X-ray diffraction (XRD) and scanning electron microscopy-energy dispersive spectroscopy (SEM-EDS) analyses indicated that Ga₃Sm and Ga₆Sm intermetallic compounds were formed on the Mo grid electrode (inert) during the potentiostatic electrolysis in LiCl-KCl-SmCl₃-AlCl₃-GaCl₃ melt, while only Ga6Sm intermetallic compound was generated on the Al-Ga alloy electrode during the galvanostatic electrolysis in LiCl-KCl-SmCl₃ melt. The electrolysis results revealed that the interaction between Sm and Ga was redominant in the Al-Ga alloy electrode, with Al only acting as an additive to lower the melting point.

Co-reduction behaviors, Sm(III), Al-Ga alloy electrode, LiCl-KCl

Radioactive materials from nuclear power facilities can be released into the atmosphere through various channels. Recently, the dispersion of radioactive materials has become critical issue in Korea after Kori Unit 1 and Wolsong Unit 1 were permanently shut down. In this study, annual atmospheric dispersion factors were compared based on the continuous release and purge release using the XOQDOQ computer program, a method for calculating atmospheric dispersion factors at commercial nuclear power stations. The meteorological data analyzed in this study was based on the Shin Kori nuclear power meteorological tower which has the largest operating nuclear power plants in Korea, for three years (from 2008 to 2010). The analysis results of the dispersion factor of the radioactive material release obtained using the XOQDOQ program showed that the difference between the continuous release and purge release was within two times. This study will be valuable helpful for revealing the uncertainty of the predictive atmospheric dispersion factor to achieve regulation.

Purge release, Continuous release, Gaseous radioactive effluent, Atmospheric dispersion factors

Chemical equilibrium calculations for multicomponent aqueous systems involving the reductive dissolution of magnetite (Fe₃O₄) with oxalic acid (H₂C₂O₄) were performed using the HSC Chemistry® version 9. They were conducted with an aqueous solution model based on the Pitzer’s approach of one molality aqueous solution. The change in the amounts and activity coefficients of species and ions involved in the reactions as well as the solution pH at equilibrium was calculated while changing the amounts of raw materials (Fe₃O₄ and H₂C₂O₄) and the system temperature from 25°C to 125°C. In particular, the conditions under which Fe₃O₄ is completely dissolved at high temperatures were determined by varying the raw amount of H₂C₂O₄ and the temperature for a given raw amount of Fe₃O₄ fed into the aqueous solution. When the raw amount of H₂C₂O₄ added was small for a given raw amount of Fe₃O₄, no undissolved Fe₃O₄ was present in the solution and the pH of the solution increased significantly. The formation of ferrous oxalate complex (FeC₂O₄) was observed. The equilibrium amount of FeC₂O₄ decreased as the raw amount of H₂C₂O₄ increased.

Chemical decontamination, Chemical equilibrium, Reductive dissolution, Magnetite, Oxalic acid

In this study, the well-known non-destructive acoustic emission (AE) and electrical resistivity methods were employed to predict quantitative damage in the silo structure of the Wolsong Low and Intermediate Level Radioactive Waste Disposal Center (WLDC), Gyeongju, South Korea. Brazilian tensile test was conducted with a fully saturated specimen with a composition identical to that of the WLDC silo concrete. Bi-axial strain gauges, AE sensors, and electrodes were attached to the surface of the specimen to monitor changes. Both the AE hit and electrical resistance values helped in the anticipation of imminent specimen failure, which was further confirmed using a strain gauge. The quantitative damage (or damage variable) was defined according to the AE hits and electrical resistance and analyzed with stress ratio variations. Approximately 75% of the damage occurred when the stress ratio exceeded 0.5. Quantitative damage from AE hits and electrical resistance showed a good correlation (R = 0.988, RMSE = 0.044). This implies that AE and electrical resistivity can be complementarily used for damage assessment of the structure. In future, damage to dry and heated specimens will be examined using AE hits and electrical resistance, and the results will be compared with those from this study.

Damage, Wolsong Low and Intermediate Level Radioactive Waste Disposal Center, Silo, Acoustic emission,Electrical resistivity

Currently, the interim storage pools of spent fuels in South Korea are expected to become saturated from 2024. It is required to prepare an operation plan of a domestic dry storage facility during a long-term period, with the researches on safety evaluation methods. This study modified the FRAPCON code to predict the spent fuel integrity evaluation such as the axial cladding temperature, the hoop stress and hydrogen distribution in dry storage. The cladding temperature in dry storage was calculated using the COBRA-SFS code with the burnup information which was calculated using the FRAPCON code. The hoop stress was calculated using the ideal gas equation with spent fuel information such as rod internal pressure. Numerical analysis method was used to calculate the degree of hydrogen diffusion according to the hydrogen concentration and temperature distribution during a dry storage period. Before 50 years of dry storage, the cladding temperature and hoop stress decreased rapidly. However, after 50 years, they decreased gradually and the cladding temperature was below 400 K. The initial temperature distribution and hydrogen concentration showed a parabolic line, but hydrogen was transferred by the hydrogen concentration and temperature gradient over time.

Dry storage, FRAPCON code, Cladding temperature, Hoop stress, Hydrogen distribution

The domestic CANDU nuclear power plants have been operated for a long time and various unforeseen spent fuel defects have been discovered. As the spent fuel defects are important factors in the safety of the nuclear power plant, a study on the analysis of the spent fuel defects to prevent their recurrence is necessary. However, in cases where the fuel rods inside the fuel assembly are defected, it is difficult to dismantle the fuel assembly owing to their welded structure and the facility conditions of the plant. Therefore, it is impossible to analyze the spent fuel defect because it is difficult to visually check the shape of the fuel defect. To resolve these problems, an analysis technology that can predict the number of defected fuel rods and defect size was developed. In this study, we developed a methodology for investigating the root cause of spent fuel defects using a database of the earlier fuel defects in the plants. It is anticipated that in the future this analysis technology will be applied when spent fuel defects occur.

CANDU, Nuclear power plant, Spent fuel defect, Defect cause, Nuclide

The reaction between Li₂CO₃ and Cl₂ was investigated to verify its occurrence during a carbon-anode-based oxide reduction (OR) process. The reaction temperature was identified as a key factor that determines the reaction rate and maximum conversion ratio. It was found that the reaction should be conducted at or above 500℃ to convert more than 90% of the Li₂CO₃ to LiCl. Experiments conducted at various total flow rate (Q) / initial sample weight (W_i) ratios revealed that the reaction rate was controlled by the Cl₂ mass transfer under the experimental conditions adopted in this work. A linear increase in the progress of reaction with an increase in Cl₂ partial pressure (pCl₂) was observed in the pCl₂ region of 2.03–10.1 kPa for a constant Q of 100 mL∙min⁻1 and W_i of 1.00 g. The results of this study indicate that the reaction between Li₂CO₃ and Cl₂ is fast at 650℃ and the reaction is feasible during the OR process.

Lithium carbonate, Chlorination reaction, Oxide reduction, Pyroprocessing, Carbon anode

A series of three bench-scale experiments was performed to investigate the conversion of sodium metal to sodium chloride via reactions with non-metal and metal chlorides. Specifically, batches of molten sodium metal were separately contacted with ammonium chloride and ferrous chloride to form sodium chloride in both cases along with iron in the latter case. Additional ferrous chloride was added to two of the three batches to form low melting point consolidated mixtures of sodium chloride and ferrous chloride, whereas consolidation of a sodium-chloride product was performed in a separate batch. Samples of the products were characterized via X-ray diffraction to identify attendant compounds. The reaction of sodium metal with metered ammonium chloride particulate feeds proceeded without reaction excursions and produced pure colorless sodium chloride. The reaction of sodium metal with ferrous chloride yielded occasional reaction excursions as evidenced by temperature spikes and fuming ferrous chloride, producing a dark salt-metal mixture. This investigation into a method for controlled conversion of sodium metal to sodium chloride is particularly applicable to sodium containing elevated levels of radioactivity—including bond sodium from nuclear fuels—in remote-handled inert-atmosphere environments.

Sodium metal, Sodium chloride, Ferrous chloride, Ammonium chloride, Sodium metal deactivation, Bond sodium

In the dismantling process of a reactor coolant system (RCS) piping, a radiation protection plan should be established to minimize the radiation exposure doses of dismantling workers. Hence, it is necessary to estimate the individual effective dose in the RCS piping dismantling process when decommissioning a nuclear power plant. In this study, the radiation exposure doses of the dismantling workers at different positions was estimated using the MicroShield dose assessment program based on the NUREG/CR-1595 report. The individual effective dose, which is the sum of the effective dose to each tissue considering the working time, was used to estimate the radiation exposure dose. The estimations of the simulation results for all RCS piping dismantling tasks satisfied the dose limits prescribed by the ICRP-60 report. In dismantling the RCS piping of the Kori-1 or Wolsong-1 units in South Korea, the estimation and reduction method for the radiation exposure dose, and the simulated results of this study can be used to implement the radiation safety for optimal dismantling by providing information on the radiation exposure doses of the dismantling workers.

Individual effective dose, RCS piping, MicroShield program, NUREG/CR-1595, ICRP-60

Solid-state mechanochemical reduction combined with subsequent melting consolidation was suggested as a technical option for the oxide reduction in pyroprocessing. Ni ingot was produced from NiO as a starting material through this technique while Li metal was used as a reducing agent. To determine the technical feasibility of this approach for pyroprocessing, which handles spent nuclear fuels, thermodynamic calculations of the phase stabilities of various metal oxides of U and other fission elements were made when several alkaline and alkali-earth metals were used as reducing agents. This technique is expected to be beneficial, not only for oxide reduction but also for other unit processes involved in pyroprocessing.

Spent nuclear fuel, Pyroprocessing, Oxide reduction, Mechanochemical reaction

Long-term experiments have been conducted on two important safety issues: long-term durability of a concrete barrier with the steel reinforcements and gas generation from low-and intermediate-level wastes in an underground research tunnel of a radioactive waste disposal facility. The gas generation and microbial communities were monitored from waste packages (200 L and 320 L) containing simulated dry active wastes. In the concrete experiment, corrosion sensors were installed on the steel reinforcements which were embedded 10 cm below the surface of concrete in a concrete mock-up, and groundwater was fed into the mock-up at a pressure of 2.1 bars to accelerate groundwater infiltration. No clear evidence was observed with respect to corrosion initiation of the steel reinforcement for 4 years of operation. This is attributed to the high integrity and low hydraulic conductivity of the concrete. In the gas generation experiment, significant levels of gas generation were not measured for 4 years. These experiments are expected to be conducted for a period of more than 10 years.

LILW, A concrete barrier, Gas generation, An underground research tunnel

In this study, a chlorination technique for recycling Li₂ZrO₃, a reaction product of ZrO₂-assisted rinsing process, was investigated to minimize the generation of secondary radioactive pyroprocessing waste. It was found that the reaction temperature was a key parameter that determined the reaction rate and maximum conversion ratio. In the temperature range of 400−600℃, an increase in the reaction temperature resulted in a profound increase in the reaction rate. Hence, according to the experimental results, a reaction temperature of at least 450℃ was proposed to ensure a Li₂ZrO₃ conversion ratio that exceeded 80% within 8 h of the reaction time. The activation energy was found to be 102 ± 2 kJ·mol⁻1·K⁻1 between 450 and 500℃. The formation of LiCl and ZrO₂ as reaction products was confirmed by X-ray diffraction analysis. The experimental results obtained at various total flow rates revealed that the overall reaction rate depends on the Cl₂ mass transfer rate in the experimental condition. The results of this study prove that the chlorination technique provides a solution to minimize the amount of radioactive waste generated during the ZrO₂-assisted rinsing process.

Chlorination reaction, Oxide reduction, Lithium zirconate, Lithium chloride, Molten salt rinsing

For application in nuclear decommissioning, underwater laser cutting studies were conducted on thick stainless-steel plates for various cutting directions using a 6 kW fiber laser. For cutting along the horizontal direction with horizontal laser irradiation, the maximum cutting speed was 110 mm∙min⁻¹ for a 48 mm thick stainless-steel plate. For cutting along the vertical direction with horizontal laser irradiation, a maximum speed of 120 mm∙min⁻¹ was obtained for the same thickness, which confirmed that the cutting performance was similar but slightly better. Moreover, when cutting with vertically downward laser irradiation, the maximum cutting speed was 120 mm∙min⁻¹ for a plate of the same thickness. Thus, the cutting performance for vertical irradiation was nearly identical to that for horizontal irradiation. In conclusion, it was possible to cut thick stainless-steel plates regardless of the laser irradiation and cutting directions, although the assist gas rose up due to buoyancy. These observations are expected to benefit laser cutting procedures during the actual dismantling of nuclear facilities.

Underwater cutting, Laser cutting, Laser processing, Nuclear dismantling, Nuclear decommissioning

A laser scabbling experiment was performed using a high-power fiber laser to investigate the removal rate of the concrete block and the scabbled depth. Concrete specimens with a 28-day compressive strength of 30 MPa were used in this study. Initially, we conducted the scabbling experiment under a stationary laser beam condition to determine the optimum scan speed. The laser interaction time with the concrete surface varied between 3 s and 40 s. The degree of spalling and vitrification on the surface was primarily dependent on the laser interaction time and beam power. Furthermore, thermal images were captured to investigate the spatial and temporal distribution of temperature during the scabbling process. Based on the experimental results, the scan speed at which the optical head moved over the concrete was set to be 300 mm∙min⁻¹ or 600 mm∙min⁻¹ for the 4.8-kW or 6.8-kW laser beam, respectively. The spalling rates and average depth on the concrete blocks were measured to be 87 cm³∙min⁻¹ or 227 cm³∙min⁻¹ and 6.9 mm or 9.8 mm with the 4.8-kW or 6.8-kW laser beams, respectively.

Laser decontamination, Laser scabbling, Spalling, Vitrification, Normal strength concrete, Thermal image

In this study, basic strategies for the decommissioning and site remediation of the Fukushima Daiichi Nuclear Power Station (FDNPS) were investigated. Six scenarios were formulated based on two of the three decommissioning strategies of nuclear power plants defined by the International Atomic Energy Agency (IAEA): immediate dismantling and deferred dismantling. A multicriteria decision analysis was performed to analyze the preferences of the options from the viewpoints of the timeframe to complete decommissioning, the resulting waste, the site usability, and the availability of the radioactive waste disposal route. The same six scenarios were applied to both the FDNPS and the nuclear power plants that ceased operation after a normal plant life cycle for comparison. For the FDNPS, the decommissioning project involved fuel debris retrieval, dismantling, and site remediation. The analysis results suggest that the balance between the amount of waste and the time to achieve the end state may be one of the most critical factors to consider when planning the decommissioning and site remediation of the FDNPS.

Fukushima Daiichi Nuclear Power Station, Site remediation, Multi-criteria decision analysis, Radioactive waste, End state, Decommissioning scenario

This study developed an analytical methodology for the mechanical integrity of spent nuclear fuel (SNF) cladding tubes under external pinch loads during transportation, with reference to the failure mode specified in the relevant guidelines. Special consideration was given to the degraded characteristics of SNF during dry storage, including oxide and hydride contents and orientations. The developed framework reflected a composite cladding model of elastic and plastic analysis approaches and correlation equations related to the mechanical parameters. The established models were employed for modeling the finite elements by coding their physical behaviors. A mechanical integrity evaluation of 14 × 14 PWR SNF was performed using this system. To ensure that the damage criteria met the applicable legal requirements, stress-strain analysis results were separated into elastic and plastic regions with the concept of strain energy, considering both normal and hypothetical accident conditions. Probabilistic procedures using Monte Carlo simulations and reliability evaluations were included. The evaluation results showed no probability of damage under the normal conditions, whereas there were small but considerably low probabilities under accident conditions. These results indicate that the proposed approach is a reliable predictor of SNF mechanical integrity.

Spent nuclear fuel, Dry storage, Mechanical integrity, Transportation, Degradation, Probability

Silver tellurite glasses with melting temperature of approximately 700°C were developed to immobilize ¹²⁹I wastes. Longterm dissolution tests in 0.1 M acetic acid and disposability assessment were conducted to evaluate sustainability of the glasses. Leaching rate of Te, Bi and I from the glasses decreased for up to 16 d, then remained stable afterwards. On the contrary, tens to tens of thousands of times more of Ag was leached in comparison to the other elements; additionally, Ag leached continuously for all 128 d of the test owing to the exchange of Ag⁺ and H⁺ ions between the glasses and solution. The I leached much lower than those of other elements even though it leached ~10 times more in 0.1 M acetic acid than in deionized water. Some TeO₄ units in the glass network were transformed to TeO₃ by ion exchange and hydrolysis. These silver tellurite glasses met all waste acceptance criteria for disposal in Korea.

Silver tellurite glass, Iodine, Long-term leaching behavior, Waste acceptance criteria

Extensive studies have been conducted on thermal conductivity of bentonite buffer materials, as it affects the safety performance of barriers engineered to contain high-level radioactive waste. Bentonite is composed of several minerals, and studies have shown that the difference in the thermal conductivity of bentonites is due to the variation in their mineral composition. However, the specific reasons contributing to the difference, especially with regard to the thermal conductivity of bentonites with similar mineral composition, have not been elucidated. Therefore, in this study, bentonites with significantly different thermal conductivities, but of similar mineral compositions, are investigated. Most bentonites contain more than 60% of montmorillonite. Therefore, it is believed that the exchangeable cations of montmorillonite could affect the thermal conductivity of bentonites. The effect of bentonite type was comparatively analyzed and was verified through the effective medium model for thermal conductivity. Our results show that Ca-type bentonites have a higher thermal conductivity than Na-type bentonites.

Bentonite buffer, Thermal conductivity, Mineral composition, Bentonite type

Legacy waste from the decommissioned A-1 nuclear power plant in the Slovak Republic is scheduled for immobilisation within a tailored alkali borosilicate glass formulation, as part of ongoing site cleanup. The aqueous durability and characterisation of a simulant glass wasteform for Chrompik III legacy waste, was investigated, including dissolution experiments up to 112 days (90°C, ASTM Type 1 water). The wasteform was an amorphous, light green glassy product, with no observed phase separation or crystalline inclusions. Aqueous leach testing revealed a suitably durable product over the timescale investigated, comparing positively to other simulant nuclear waste glasses and vitreous products tested under similar conditions. Iron and titanium rich precipitates were observed to form at the surface of monolithic samples during leaching, with the formation of an alkali deficient alteration layer behind these at later ages. Overall this glass appears to perform well, and in line with expectations for this chemistry, although longer-term testing would be required to predict overall durability. This work will contribute to developing confidence in the disposability of vitrified Chrompik legacy wastes.

Higher activity waste, Vitrification, Glass, Durability, Characterisation, Chrompik

The solubilities of different multicomponent lanthanide oxide (Ln₂O₃) solid solutions including binary (Ln₁ and Ln₂ = La, Nd, Eu, or Tm), ternary (Ln₁, Ln₂, and Ln₃ = La, Nd, Eu, or Tm), and higher systems (Ln = La, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, and Lu) were studied after aging for four weeks at 60°C. Our recent study revealed that the phase transformations in binary ((La, Nd) and (La, Eu)) and ternary (La, Nd, Eu) systems are responsible for the formation of (La, Nd)(OH)₃, (La, Eu)(OH)₃, and (La, Nd, Eu)(OH)₃ solid solutions, respectively. The variations in the mole fractions of La³⁺, Nd³⁺, and Eu³⁺ in the sample solutions of these hydroxide solid solutions indicated that a thermodynamic equilibrium might account for the apparent La, Nd, and Eu solubilities. Conversely, the binary and ternary systems containing Tm₂O₃ as the heavy lanthanide oxide retained the oxide-based solid solutions, and their solubility behaviors were dominated by their congruent dissolutions. In the higher multicomponent system, the X-ray diffraction patterns of the solid phases, before and after contact with the aqueous phase indicated the formation of a stable oxide solid solution and their solubility behavior was explained by its congruent dissolution.

Lanthanide, Oxide Hydroxide, Solid solution, Solubility

The overestimation and underestimation of the radioactivity concentration of difficult-to-measure radionuclides can occur during the implementation of the scaling factor (SF) method because of the uncertainties associated with sampling, radiochemical analysis, and application of SFs. Strict regulations ensure that the SF method as an indirect method does not underestimate the radioactivity of nuclear wastes; however, there are no clear regulatory guidelines regarding the overestimation. This has been leading to the misuse of the SF methodology by stakeholders such as waste disposal licensees and regulatory bodies. Previous studies have reported instances of overestimation in statistical implementation of the SF methodology. The analysis of the two most popular linear models of the SF methodology showed that severe overestimation may occur and radioactivity concentration data must be dealt with care. Since one major source of overestimation is the use of minimum detectable activity (MDA) values as true activity values, a comparative study of instrumental techniques that could reduce the MDAs was also conducted. Thermal ionization mass spectrometry was recommended as a suitable candidate for the trace level analysis of long-lived beta-emitters such as iodine-129. Additionally, the current status of the United States and Korea was reviewed from the perspective of overestimation.

Scaling factor, Overestimation, Radioactivity concentration, Minimum detectable activity

The challenges facing companies and institutions surrounding civil nuclear decommissioning are diverse and many, none more so than those faced in the United Kingdom. The UK’s Generation I nuclear power plants and early research facilities have left a ‘Nuclear Legacy’ which is in urgent need of management and clean-up. Sellafield is quite possibly the most illfamed nuclear site in the UK. This complex and challenging site houses much of what is left from the early days of nuclear research in the UK, including early nuclear reactors (Windscale Piles, Calder Hall, and the Windscale Advanced Gas Cooled Reactor) and the UK’s early nuclear weapons programme. Such a legacy now requires careful management and planning to safely deal with it. This task falls on the shoulders of the Nuclear Decommissioning Authority (NDA). Through a mix of prompt and delayed decommissioning strategies, key developments in R&D, and the implementation of site licenced companies to enact decommissioning activities, the NDA aims to safety, and in a timely manner, deal with the UK’s nuclear legacy. Such approaches have the potential to influence and shape other such approaches to nuclear decommissioning activities globally, including in Korea.

Nuclear Decommissioning, Nuclear Decommissioning Authority, Sellafield, Nuclear legacy, United Kingdom, Korea

The present multidisciplinary study, which is a nexus of engineering and political science, investigates how the modernization of Non-Strategic Nuclear Weapons (NSNWs) affects the IAEA safeguards system based on the likelihood of the use of nuclear weapons. To this end, this study examines the characteristics of modernized NSNWs using Monte Carlo techniques. The results thus obtained show that 10 kt NSNWs with a Circular Error Probability (CEP) of 10 m can destroy the target as effectively as a 500 kt weapon with a CEP of 100 m. The IAEA safeguards system shows that the Significant Quantity (SQ) of 1 of plutonium is 8 kg, a parameter that was established when strategic nuclear weapons were dominant. However, the results of this study indicate that in recent years, low-yield nuclear weapons such as NSNWs have been more strategically interesting than strategic nuclear weapons as NSNWs require less plutonium than strategic nuclear weapons. Therefore, we would like to conclude that reducing the SQ of plutonium can result in more robust safeguards and non-proliferation strategies.

Non-strategic nuclear weapon, Low-yield nuclear weapon, Monte Carlo simulation, Nuclear Safeguards, Significant quantity, Plutonium

This article examines the consequences of a significant spent fuel management decision or event in the United States, South Korea and Taiwan. For the United States, it is the financial impact of the Department of Energy’s inability to take possession of spent fuel from commercial nuclear power companies beginning in 1998 as directed by Congress. For South Korea, it is the potential financial and socioeconomic impact of the successful construction, licensing and operation of a low and intermediate level waste disposal facility on the siting of a spent fuel/high level waste repository. For Taiwan, it is the operational impact of the Kuosheng 1 reactor running out of space in its spent fuel pool. From these, it draws six broad lessons other countries new to, or preparing for, nuclear energy production might take from these experiences. These include conservative planning, treating the back-end of the fuel cycle holistically and building trust through a step-by-step approach to waste disposal.

Disposal, Storage, Spent fuel, Dry cask, Repository

Graphite Isotope Ratio Method (GIRM) can be used to estimate plutonium production in a graphite-moderated reactor. This study presents verification results for the GIRM combined with a 3-D polynomial regression function to estimate cumulative plutonium production in a graphite-moderated reactor. Using the 3-D Monte-Carlo method, verification was done by comparing the cumulative plutonium production with the GIRM. The GIRM can estimate plutonium production for specific sampling points using a function that is based on an isotope ratio of impurity elements. In this study, the ¹⁰B/¹¹B isotope ratio was chosen and calculated for sampling points. Then, 3-D polynomial regression was used to derive a function that represents a whole core cumulative plutonium production map. To verify the accuracy of the GIRM with polynomial regression, the reference value of plutonium production was calculated using a Monte-Carlo code, MCS, up to 4250 days of depletion. Moreover, the amount of plutonium produced in certain axial layers and fuel pins at 1250, 2250, and 3250 days of depletion was obtained and used for additional verification. As a result, the difference in the total cumulative plutonium production based on the MCS and GIRM results was found below 3.1% with regard to the root mean square (RMS) error.

Graphite Isotope Ratio Method, GIRM, MCS, Plutonium production, Non-proliferation

In the majority of countries, the upper limit of buffer temperature in a repository is set to below 100℃ due to the possible illitization. This smectite-to-illite transformation is expected to be detrimental to the swelling functions of the buffer. However, if the upper limit is increased while preventing illitization, the disposal density and cost-effectiveness for the repository will dramatically increase. Thus, understanding the characteristics and creating a database related to the buffer under the elevated temperature conditions is crucial. In this study, a strategy to investigate the bentonite found in Korea under the elevated temperatures from a mineral transformation and radionuclides retardation perspective was proposed. Certain long-term hydrothermal reactions generated the bentonite samples that were utilized for the investigation of their mineral transformation and radionuclide retardation characteristics. The bentonite samples are expected to be studied using in-situ synchrotron-based X-Ray Diffraction (XRD) technique to determine the smectite-to-illite transformation. Simultaneously, the ‘high-temperature and high-pressure mineral alteration measurement system’ based on the Diamond Anvil Cell (DAC) will control and provide the elevated temperature and pressure conditions during the measurements. The kinetic models, including the Huang and Cuadros model, are expected to predict the time and manner in which the illitization will become detrimental to the performance and safety of the repository. The sorption reactions planned for the bentonite samples to evaluate the effects on retardation will provide the information required to expand the current knowledge of repository optimization.

Korean bentonite, Elevated temperature conditions, Smectite-to-illite transformation, Radwaste disposal

Safe storage of spent nuclear fuel in deep underground repositories necessitates an understanding of the long-term alteration of metal canisters and buffer materials. A small-scale laboratory alteration test was performed on metal (Cu or Fe) chips em!bedded in compacted bentonite blocks placed in anaerobic water for 1 year. Lactate, sulfate, and bacteria were separately added to the water to promote biochemical reactions in the system. The bentonite blocks immersed in the water were dismantled after 1 year, showing that their alteration was insignificant. However, the Cu chip exhibited some microscopic etch pits on its surface, wherein a slight sulfur component was detected. Overall, the Fe chip was more corroded than the Cu chip under the same conditions. The secondary phase of the Fe chip was locally found as carbonate materials, such as siderite (FeCO₃) and calcite ((Ca, Fe)CO₃). These secondary products can imply that the local carbonate occurrence on the Fe chip may be initiated and developed by an evolution (alteration) of bentonite and a diffusive provision of biogenic CO₂ gas. These laboratory scale results suggest that the actual long-term alteration of metal canisters/bentonite blocks in the engineered barrier could be possible by microbial activities.

Repositories, Alteration, Canisters, Bentonite, Bacteria

The Republic of Korea is expected to participate in the denuclearization verification activities by the International Atomic Energy Agency (IAEA) in case any neighboring countries declared denuclearization. In this study, samples for the verification of nuclear activities in undeclared areas were selected for the denuclearization of neighboring countries, and the appropriateness of the procedures was considered. If a country with nuclear weapons declares denuclearization, it must be accompanied by the IAEA’s verification regarding nuclear materials and weapons in the declared and undeclared areas. The analysis of the process samples or on-site environmental samples and the verification of undeclared nuclear facilities and materials aid in uncovering any evidence of concealment of nuclear activity in undeclared areas. Therefore, a methodology was established for effective sampling and analysis in accordance with proper procedures. Preparations for sampling in undeclared areas were undertaken for various potential scenarios, such as, the establishment of zones according to radiation dose, methods of supplying electricity, wireless communication networks, targets of sampling according to characteristics of nuclides, manned sampling method, and unmanned sampling method. Through this, procedures were established for pre- and post-site settings in preparation for hazards and limiting factors at nuclear inspection sites.

Denuclearization, Verification, Sampling, Soil, Swipe

This paper aims to evaluate the mechanical integrity for Spent Nuclear Fuel (SNF) cladding under lateral loads during transportation. The evaluation process requires a conservative consideration of the degradation conditions of SNF cladding, especially the hydride effect, which reduces the ductility of the cladding. The dynamic forces occurring during the drop event are pinch force, axial force and bending moment. Among those forces, axial force and bending moment can induce transverse tearing of cladding. Our assessment of 14 × 14 PWR SNF was performed using finite element analysis considering SNF characteristics. We also considered the probabilistic procedures with a Monte Carlo method and a reliability evaluation. The evaluation results revealed that there was no probability of damage under normal conditions, and that under accident conditions the probability was small for transverse failure mode.

Spent nuclear fuel, Dry storage, Mechanical integrity, Transportation, Transverse tearing, Probability

In this study, rainfall infiltration in vault of the second near-surface disposal facility was evaluated on the basis of various disposal scenarios. A total of four different disposal scenarios were examined based on the locations of the radioactive waste containers. A numerical model was developed using the FEFLOW software and finite element method to simulate the behavior of infiltrated water in each disposal scenario. The effects of the disposal scenarios on the infiltrated water were evaluated by estimating the flux of the infiltrated water at the vault interfaces. For 300 years, the flux of infiltrated water flowing into the vault was estimated to be 1 mm/year or less for all scenario. The overall results suggest that when the engineered barriers are intact, the flux of infiltrated water cannot generate a sufficient pressure head to penetrate the vault. In addition, it is confirmed that the disposal scenarios have insignificant effects on the infiltrated water flowing into the vault.

Rainfall infiltration, Near-surface disposal facility, Vault, Emplacement of radioactive waste, Flux of infiltrated water

Safe geological disposal of spent nuclear fuel (SNF) requires knowledge of the deep hydrochemical characteristics of the repository site. Here, we conducted a set of deep hydrochemical investigations using a 750-m borehole drilled in a model granite system in Wonju, South Korea. A closed investigation system consisting of a double-packer, Waterra pump, flow cell, and water-quality measurement unit was used for in situ water quality measurements and subsequent groundwater sampling. We managed the drilling water labeled with a fluorescein dye using a recycling system that reuses the water discharged from the borehole. We selected the test depths based on the dye concentrations, outflow water quality parameters, borehole logging, and visual inspection of the rock cores. The groundwater pumped up to the surface flowed into the flow cell, where the in situ water quality parameters were measured, and it was then collected for further laboratory measurements. Atmospheric contact was minimized during the entire process. Before hydrochemical measurements and sample collection, pumping was performed to purge the remnant drilling water. This study on a model borehole can serve as a reference for the future development of deep hydrochemical investigation procedures and techniques for siting processes of SNF repositories.

SNF, Geological disposal, Hydrochemical investigation, Granite, Deep borehole

Large earthquakes with (M_W > ~ 6) result in ground shaking, surface ruptures, and permanent deformation with displacement. The earthquakes would damage important facilities and infrastructure such as large industrial establishments, nuclear power plants, and waste disposal sites. In particular, earthquake ruptures associated with large earthquakes can affect geological and engineered barriers such as deep geological repositories that are used for storing hazardous radioactive wastes. Earthquake-driven faults and surface ruptures exhibit various fault zone structural characteristics such as direction of earthquake propagation and rupture and asymmetric displacement patterns. Therefore, estimating the respect distances and hazardous areas has been challenging. We propose that considering multiple parameters, such as fault types, distribution, scale, activity, linkage patterns, damage zones, and respect distances, enable accurate identification of the sites for deep geological repositories and important facilities. This information would enable earthquake hazard assessment and lower earthquakeresulted hazards in potential earthquake-prone areas.

Fault damage zone, Active fault, Surface rupture, Respect distance, Deep geological disposal

Molten salt solutions consisting of eutectic LiCl-KCl and concentrations of samarium chloride (0.5 to 3.0 wt%) at 500℃ were analyzed using both cyclic voltammetry (CV) and electrochemical impedance spectroscopy (EIS). The CV technique gave the average diffusion coefficient for Sm³⁺ over the concentration range. Equipped with Sm³⁺ diffusion coefficient, the Randles-Sevcik equation predicted Sm³⁺ concentration values that agree with the given experimental values. From CV measurements; the anodic, cathodic, and half-peak potentials were identified and subsequently used as a parameter to acquire EIS spectra. A six-element Voigt model was used to model the EIS data in terms of resistance-time constant pairs. The lowest resistances were observed at the half-peak potential with the associated resistance-time constant pairs characterizing the reversible reaction between Sm³⁺ and Sm²⁺. By extrapolation, the Voigt model estimated the polarization resistance and established a polarization resistance-concentration relationship.

Molten salt,Electrochemical impedance spectroscopy,Cyclic voltammetry,Concentration,Samarium chloride

This paper gives two graph-based algorithms for radioactive decay computation. The first algorithm identifies the connected components of the graph induced from the given radioactive decay dynamics to reduce the size of the problem. The solutions are derived over the precalculated connected components, respectively and independently. The second algorithm utilizes acyclic structure of radioactive decay dynamics. The algorithm evaluates the reachable vertices of the induced system graph from the initially activated vertices and finds the minimal set of starting vertices populating the entire reachable vertices. Then, the decay calculations are performed over the reachable vertices from the identified minimal starting vertices, respectively, with the partitioned initial value over the reachable vertices. Formal arguments are given to show that the proposed graph inspired divide and conquer calculation methods perform the intended radioactive decay calculation. Empirical efforts comparing the proposed radioactive decay calculation algorithms are presented.

Radioactive decay computation,Nuclear fuel cycle,Graph algorithms

We studied the anisotropic shrinkage and deformation characteristics of large size sintered bodies in the manufacturing of glass-ceramic composite wasteform. We used uranium-bearing waste, generated from the treatment of spent uranium catalyst. Sintered specimens were prepared in several forms, comprising a circular disk, and a quarter disk in several diameters of up to 40 cm. Regardless of form or size, the sintered bodies had high isotropic shrinkage when they were fabricated using green bodies prepared at 60 MPa. The average anisotropy rate and average shrinkage rate were 1.6%, and 37.4%, respectively. We confirmed that the glass-ceramic composite wasteform in a large scale disk-type for packing in a 200 L drum could be fabricated with a tolerable anisotropy shrinkage. This has resulted in a significant reduction in the volume of radioactive waste to be disposed of with highly stable wasteform.

Radioactive wasteform,Isotropy shrinkage,Immobilization,Glass-ceramic composite,Volume reduction,Uranium catalyst

An engineered barrier system (EBS) for the geological disposal of high-level radioactive waste (HLW) consists of a disposal canister packed with spent fuel, buffer material, backfill material, and gap-filling material. The buffer material fills the space between the canister and the near-field rock, thus serving to restrain the release of radionuclides and protect the canister from groundwater penetration. Furthermore, as significant amounts of heat energy are released from the canister to the surrounding rock, the thermal conductivity of the buffer plays an important role in maintaining the safety of the entire disposal system. Therefore, given the high levels of heat released from disposal canisters, this study measured the thermal conductivities of compacted bentonite buffers from Gyeongju under temperature variations ranging 25 to 80~90℃. There was a 5~20% increase in thermal conductivity as the temperature increased, and the temperature effect increased as the degree of saturation increased.

Engineered barrier system,Compacted bentonite buffer,Thermal conductivity

This study was conducted to develop a method for depth assessment of embedded sources using gamma-spectrum ratio and for the evaluation of field applicability. To this end, Peak to Compton and Peak to valley ratio changes were evaluated according to ¹³⁷Cs, ⁶⁰Co, ¹⁵²Eu point source depth using HPGe detector and MCNP simulation. The effects of measurement distance of PTV and PTC methods were evaluated. Using the results, the source depth assessment equation using the PTC and PTV methods was derived based on the detection distance of 50 cm. In addition, the sensitivity of detection distance changes was assessed when using PTV and PTC methods, and error increased by 3 to 4 cm when detection distance decreased by 20 cm based on 50 cm. However, it was confirmed that if the detection distance was increased to 100 cm, the effects of detection distance were small. And PTV and PTC methods were compared with the two distance measurement method which evaluates the depth of source by the change of net peak counting rate according to the detection distance. As a result of source depth assessment, the PTV and PTC showed a maximum error of 1.87 cm and the two distance measurement method showed maximum error of 2.69 cm. The results of the experiment confirmed that the accuracy of the PTV and PTC methods was higher than two distance measurement. In addition, Sensitivity evaluation by horizontal position error of source has maximum error of less than 25.59 cm for the two distance measurement method. On the other hand, PTV and PTC method showed high accuracy with maximum error of less than 8.04 cm. In addition, the PTC method has lowest standard deviation for the same time measurement, which is expected to enable rapid measurement.

HPGe detector,In-situ measurement,Hotspot,Source depth,Peak to Compton,Peak to Valley

Dissolved hexavalent uranium can exist in the form of several different chemical species. Furthermore, species distributions depend on the pH value of the aqueous solution. Representatively, UO₂²⁺, UO₂OH⁺, (UO₂)₂(OH)₂²⁺, and (UO_{2)3(OH)5⁺ species coexist in solutions at acidic and circumneutral pH values. When amorphous silica particles are suspended in an aqueous solution, the dissolved chemical species are easily adsorbed onto silica surfaces. In this study, it was examined whether the species distribution of the adsorbed U(VI) on a silica surface followed that of the dissolved U(VI) in an aqueous solution. Time-resolved luminescence spectra of three different dissolved species(UO2²⁺, UO2OH⁺, and (UO2)3(OH)5⁺) and two different adsorbed species (≡SiO2UO2}, ≡SiO₂(UO₂)OH^–, or ≡SiO₂(UO₂)₃(OH)^5–) were measured in the pH range 3.5−7.5. The spectral shapes of these chemical species were compared by changing the pH value; consequently, it was confirmed that the species distribution of the adsorbed U(VI) species was different from that of the dissolved U(VI) species.

Luminescence,Hexavalent uranium,Dissolved uranium,Adsorbed uranium,Silica

Strontium (⁹⁰Sr) and nickel (⁵⁹Ni) have been considered as key radionuclides in the safety assessment of radioactive waste disposal. Through various efforts to impede the migration of radioactive nuclides underground, it has been established that some minerals generated from the corrosion of the waste containers have a positive chemical interaction with these radionuclides. Among these minerals we selected mackinawite (FeS), an iron and sulfur compound, and performed a sorption experiment for the Sr and Ni in FeS under anoxic and alkaline conditions by reflecting deep underground environments. The effects of pH on sorption were likewise investigated in the pH range of 8 ~ 12. As a result, it was found that strontium failed to exhibit a good sorption capacity in a weak alkaline range, while nickel showed a noticeably higher sorption affinity over the entire experimental pH range. Moreover, we determined that as the pH increased in the solution, the distribution coefficients (K_d) were increased for both nuclides, which reflects when an alkalinity increses, the surface of the mineral charges much negatively by detaching the hydrogen or cations on the mineral surface. Thus, it can be concluded that the cationic nuclides of Sr and Ni can attach easily to the mineral under strong alkalinity.

Strontium,Nickel,Sorption,Mackinawite,pH dependancy,Chemical extraction

The Graphite Isotope Ratio Method (GIRM) can verify non-proliferation of nuclear weapon by estimating the total plutonium production in a graphite-moderated reactor. Using the reactor, plutonium is generated and accumulated through the 238U neutron capture reaction, and impurities in the graphite are converted to nuclides due to the nuclear reaction. Therefore, the amount of plutonium production and concentration of the impurities are correlated. However, the plutonium production cannot be predicted using only the absolute concentration of the impurities. It can only be predicted when the initial concentration of the impurities is obtained because the concentration, at a certain time, depends on it. Nevertheless, the ratios of the isotopes in an element are known regardless of the impurity of an element in the graphite moderator. Thus, the correlation between the isotope ratio and amount of plutonium produced helps predict plutonium production in a graphite-moderated reactor. Boron, Lithium, Chlorine, Titanium, and Uranium are known as indicator elements in the GIRM. To assess whether the correlation between the indicator isotope and amount of plutonium produced is independent of the initial concentration of the impurities, four different impurity compositions of graphite were used. 10B/11B, 36Cl/35Cl, 48Ti/49Ti, and 235U/238U had a consistent correlation with the cumulative plutonium production, regardless of the initial impurity concentration of the graphite, because these isotopes were not generated through the nuclear reaction of other elements. On the other hand, the correlation between 6Li/7Li and plutonium production depended on the initial concentration of the impurities in graphite. Although 7Li can be produced through the neutron capture reaction of 6Li, the (n, α) reaction of 10B was the major source of 7Li. Therefore, the initial concentration of 10B affected the production of 7Li, making Li unsuitable as an indicator element for the GIRM.

Indicator isotope,Graphite Isotope Ratio Method (GIRM),Prediction of plutonium production,Graphitemoderated reactor,Non-proliferation

According to the Korea Radioactive Waste Agency’s (KORAD’s) medium and low level radioactive waste management implementation plan, the Domestic 3rd Step Landfill Disposal Facility has planned to accept a total of 104,000 drums (2 trenches) of very low level radioactive waste (VLLW), from the decommissioning site from April 2019 - February 2026 (total budget: 224.6 billion Won). Subsequently, 260,000 drums (5 trenches) will be disposed in a 34,076 m² . Accordingly, KORAD is preparing a waste acceptance criteria (WAC) for this facility. Every disposal facility for VLLW in other countries such as France and Spain, operate their WAC for each VLLW facility with a reasonable application approach, This, paper focuses on analyzing the WAC conditions in VLLW sites in the USA and discusses whether these can be met in domestic VLLW WAC. It also helps in the preparation of WAC for the 3rd Step Landfill Disposal Site in Gyeongju, since the USA has prior experience on decommissioning nuclear waste.

Decommissioning waste,Very low level radioactive waste,Acceptance criteria,Landfill disposal

Pulling-type cutting devices, which use a diamond wire saw, have been used generally for cutting concrete structures. In this study, a pushing-type cutting device with a collection cover was developed by overcoming the disadvantages of pulling-type devices. In this device, dry or liquid methods can be selected to cool frictional heat. Operation and leakage tests of the dust generated during the dismantling of a concrete structure were carried out, confirming the suitable operation of the fabricated cutting device; the leakage rate was approximately 1.7%. For a conservative evaluation, the internal dose of workers was estimated in dismantling the core center part of biological shield concrete with a specific activity of 99.5 Bq·g^-1. The committed effective dose per worker was 0.25 mSv. The developed cutting device contributed to reducing radioactive concrete waste and minimizing worker exposure due to its easy installation. Therefore, it can be utilized as a cutting apparatus for dismantling not only reinforced concrete structures but also radioactive biological shield concrete in nuclear power plant decommissioning efforts.

Pushing type,Cutting device,Operation,Leakage,Committed effective dose

An options study was performed for the treatment of residual elemental sodium in driver plenums following the chopping operation during the pyroprocessing of used nuclear fuel. Given the pending availability of a multi-function furnace for distillation and consolidation operations in the Fuel Conditioning Facility, the furnace was considered for the processing of driver plenums. Although two options (oxidation and distillation) could be performed in the multi-function furnace, neither option has been developed sufficiently to date to warrant the use of the furnace for treatment operations. Thus, it was decided to defer the treatment of elemental sodium from driver plenums in the multi-function furnace until more developed technologies and/or furnaces become available. In the interim, storage of the plenums and characterization efforts are recommended.

EBR-Ⅱ,Sodium,Pyroprocessing

A numerical model was developed using COMSOL Multiphysics to evaluate groundwater flow that causes radionuclide migration in the unsaturated zone of a near-surface disposal facility, which is considered as a domestic low and an intermediate-level radioactive waste disposal facility. Each scenario was modeled by constructing a two-dimensional domain that included the disposal vault, backfill, disposal cover, and unsaturated aquifer. A comparison of the continuous and intermittent rainfall conditions exhibited no significant difference in any of the factors considered except the wave pattern of water saturation. The input data, such as porosity and residual water content of the unsaturated aquifer, were observed to not have a significant effect on the groundwater flow. However, the hydraulic conductivity of the unsaturated aquifer was found to have a significant effect on the groundwater flow. Therefore, it is necessary to assess the hydraulic conductivity of an unsaturated aquifer to determine the extent of groundwater infiltration into the disposal vault.

Near-surface disposal system,Unsaturated aquifer,Groundwater flow,Numerical model

A geological repository comprises a natural barrier and an engineered barrier system. Its design components consist of canisters, buffers, backfill, and near-field rock. Among the engineered barrier system components, bentonite buffers minimize the groundwater flow from near-field rock and prevent the release of nuclide. Investigation of the hydraulic conductivity of the buffer to groundwater flow is an important factor in the performance evaluation of the stability and integrity of the engineered barrier of the repository. In this study, saturated hydraulic conductivity tests were performed using Gyeongju bentonite at various dry densities and temperatures, and a hydraulic conductivity prediction model was developed through multiple regression analysis using the 120 result sets of hydraulic conductivity. The test results showed that the hydraulic conductivity tends to decrease as the dry density increases. In addition, the hydraulic conductivity increased with increasing temperature. The multiple regression analysis results showed that the coefficient of determination (R²) of the hydraulic conductivity prediction equation was as high as 0.93. The hydraulic conductivity prediction equation presented in this study could be used for the design of engineered barrier systems.

Engineered barrier system,Bentonite buffer,Hydraulic conductivity,Multiple regression analysis

A reference electrode is important for controlling electrochemical reactions. Evaluating properties such as the reduction potential of the elements is necessary to optimize the electrochemical processes in pyroprocessing, especially in a multicomponent environment. In molten chloride systems, which are widely used in pyroprocessing, a reference electrode is made by enclosing the silver wire and molten salt solution containing silver chloride into the membranes. However, owing to the high temperature of the molten salt, the choice of the membrane for the reference electrode is limited. In this study, three types of electroceramic, mullite, Pyrex, and quartz, were compared as reference electrode membranes. They are widely used in molten salt electrochemical processes. The potential measurements between the two reference electrode systems showed that the mullite membrane has potential deviations of approximately 50 mV or less at temperatures higher than 650℃, Pyrex at temperatures lower than 500℃, and quartz at temperatures higher than 800℃. Cyclic voltammograms with different membranes showed a significant potential shift when different membranes were utilized. This research demonstrated the uncertainties of potential measurement by a single membrane and the potential shift that occurs because of the use of different membranes.

Mullite,Pyrex,Quartz,Open circuit potential,Mechanical properties

Removing radioactive concrete is crucial in the decommissioning of nuclear power plants. However, this process generates radioactive aerosols, exposing workers to radiation. Although large amounts of radioactive concrete are generated during decommissioning, studies on the internal exposure of workers to radioactive aerosols generated from the cutting of radioactive concrete are very limited. In this study, therefore, we calculate the internal radiation doses of workers exposed to radioactive aerosols during activities such as drilling and cutting of radioactive concrete, using previous research data. The electrical-mobility-equivalent diameter measured in a previous study was converted to aerodynamic diameter using the Newton–Raphson method. Furthermore, the specific activity of each nuclide in radioactive concrete 10 years after nuclear power plants are shut down was calculated using the ORIGEN code. Eventually, we calculated the committed effective dose for each nuclide using the IMBA software. The maximum effective dose of ¹⁵²Eu constituted 83.09% of the total dose; moreover, the five highest-ranked elements (¹⁵²Eu, ¹⁵⁴Eu, ⁶⁰Co, ²³⁹Pu, ⁵⁵Fe) constituted 99.63%. Therefore, we postulate that these major elements could be measured first for rapid radiation exposure management of workers involved in decommissioning of nuclear power plants, even if all radioactive elements in concrete are not considered.

Radioactive aerosol,Concrete,Internal exposure,AMAD,IMBA

Here, the stability of stainless steel 316 (SS-316) was investigated to identify its applicability in the oxide reduction process, as a component in related equipment, to produce a complicated gas mixture composed of O₂ and Cl₂ under an argon (Ar) atmosphere. The effects of the mixed gas composition were investigated at flow rates of 30 mL/min O₂, 20 mL/min O₂ + 10 mL/min Cl₂, 10 mL/min O₂ + 20 mL/min Cl₂, and 30 mL/min Cl₂, each at 600℃, during a constant argon flow rate of 170 mL/min. It was found that the corrosion of SS-316 by the chlorine gas was suppressed by the presence of oxygen, while the reaction proceeded linearly with the reaction time regardless of gas composition. Surface observation results revealed an uneven surface with circular pits in the samples that were fed mixed gases. Thermodynamic calculations proposed the combination of Fe and Ni chlorination reactions as an explanation for this pit formation phenomenon. An exponential increase in the corrosion rate was observed with an increase in the reaction temperature in a range of 300 ~ 600℃ under a flow of 30 mL/min Cl₂ + 170 mL/min Ar.

Stainless steel 316,Chlorine,Oxygen-chlorine mixed gas,Oxide reduction,Pyroprocessing

A Bayesian approach was introduced to improve the belief of prior distributions of input parameters for the probabilistic safety assessment of radioactive waste repository. A GoldSim-based module was developed using the Markov chain Monte Carlo algorithm and implemented through GSTSPA (GoldSim Total System Performance Assessment), a GoldSim template for generic/site-specific safety assessment of the radioactive repository system. In this study, sequential Bayesian updating of prior distributions was comprehensively explained and used as a basis to conduct a reliable safety assessment of the repository. The prior distribution to three sequential posterior distributions for several selected parameters associated with nuclide transport in the fractured rock medium was updated with assumed likelihood functions. The process was demonstrated through a probabilistic safety assessment of the conceptual repository for illustrative purposes. Through this study, it was shown that insufficient observed data could enhance the belief of prior distributions for input parameter values commonly available, which are usually uncertain. This is particularly applicable for nuclide behavior in and around the repository system, which typically exhibited a long time span and wide modeling domain.

Probabilistic safety assessment,HLW repository,Bayesian updating,Prior/posterior distribution,GoldSim,GSTSPA

Although the national radiological emergency response system has been improved by incorporating lessons from the Fukushima nuclear power plant accident and recent domestic natural disasters, it has not fully incorporated these lessons. In addition, it cannot deal with a variety of aftermath of the radiological disaster. Even for the same disaster, the national emergency response system should comply with multiple domestic laws in our country. Furthermore, there are a few discrepancies between the articles of the domestic laws that the national radiological emergency response system should address. Therefore, this study investigates the characteristics of radiological disasters, examines articles on the domestic laws related to the national radiological emergency response system, and analyses the Japanese government’s responses to the Fukushima nuclear power plant accident. Based on the results of the review, suggestions for the improvement of the national radiological emergency response system in terms of response organization and framework have been proposed in this study.

Radiological disaster,Radiological emergency response system,Fukushima nuclear power plant accident,Suggestions

The feasibility of rare earth (RE) removal process via oxidation reactions with UCl₃ was investigated using the HSC Chemistry code to reduce the concentrations of RE in transuranic (TRU) products. The composition and thermodynamic data of TRU and RE elements contained in the reference spent fuel were examined. The reactivity was evaluated by calculating equilibrium data considering oxidation reactions with UCl₃. Both RE removal rate and TRU recovery rate were evaluated for the two cases, wherein TRU products with different RE concentrations were used. When TRU products were reacted with UCl₃, selective oxidation was driven by the difference in the Gibbs free energy of each element. The calculation results imply that the TRU/RE ratio of the final product can be increased by removing RE elements while maintaining the maximum recovery rate of TRU, which is accomplished by controlling the amount of UCl₃ injected. Since the results of this study are based on thermodynamic equilibrium data, there are many limitations to apply to the actual process. However, it is expected to be used as an important data for the process design to supply the TRU product of pyroprocessing to SFR’s fuel demanding low RE concentrations.

Pyroprocessing,TRU product,RE removal by UCl₃,HSC Chemistry

Concrete waste accounts for approximately 70~80% of the total waste generated during the decommissioning of nuclear power plants (NPPs). Based upon the concentration of each radionuclide, the concrete waste from the decommissioning can be used in the determination of the clearance threshold used to classify waste as radioactive. To reduce the cost of radioactive concrete waste disposal, it is important to perform decontamination before self-disposal or limited recycling. Therefore, it is necessary to estimate the internal radioactivity distribution of radioactive concrete waste to ensure effective decontamination. In this study, the performance metrics of various Compton reconstruction algorithms were compared in order to identify the best strategy to estimate the internal radioactivity distribution in concrete waste during the decommissioning of NPPs. Four reconstruction algorithms, namely, simple back-projection, iltered back-projection, maximum likelihood expectation maximization (MLEM), and energy-deconvolution MLEM (E-MLEM) were used as Compton reconstruction algorithms. Subsequently, the results obtained by using these various reconstruction algorithms were compared with one another and evaluated, using quantitative evaluation methods. The MLEM and E-MLEM reconstruction algorithms exhibited the best performance in maintaining a high image resolution and signal-to-noise ratio (SNR), respectively. The results of this study demonstrate the feasibility of using Compton images in the estimation of the internal radioactive distribution of concrete during the decommissioning of NPPs.

Concrete waste,Compton reconstruction algorithms,CdZnTe (CZT),Decommissioning

In this study, we discussed the limitations of gross alpha measurements for the characterization of radioactive wastes produced in nuclear facilities through experimental tests and Monte Carlo N-particle transport simulations. The determination of gross alpha is essential for the disposal of radioactive waste produced in nuclear facilities in Korea. The measurements of gross alpha are easy to perform and yield rapid analytical results, but it cannot be used for quantitative analysis. The error of counting efficiency for gross alpha with various masses of the deposit on planchets using KCl and ²⁴¹Am was determined. The relative deviation of the counting efficiency in samples having the same mass was 20%. Uranium was extracted from the soil through acid leaching and extraction chromatography, and the concentration of U determined by inductively coupled plasma–mass spectrometry (ICP-MS) was compared with the results for gross alpha. The gross alpha was underestimated by 50% compared to the U concentration by ICP-MS. The counting efficiency depended on the energy from the alpha emitters, which differed by up to three times in determination of the counting efficiency depending on the kinds of alpha radionuclides of interest. Therefore, the gross alpha is not compatible with the sum of radioactivity for each alpha emitter and is suitable as a screening method.

Gross alpha,Disposal,Radioactive waste,Radiological characterization

Metal seals are the main components that establish the containment boundary in bolted casks, which store spent nuclear fuel. These seals are degraded by heat and radiation. In addition, creep occurs when the seals are exposed to intense heat for an extended period. This creep results in the stress relaxation of the seals, which consequently impairs the seal integrity. The stress relaxation can reduce the sealing performance of the metal seal, which can further cause leakage in the storage cask. Moreover, the reduction of bolt tension leads to sealing performance degradation. In this study, the results of high-temperature-accelerated tests were obtained to evaluate the containment integrity of metal seals and the decrease in bolt tension. During the tests, the leakage rate, bolt strain, and ambient temperature of the metal seals were measured and analyzed. The metal seals were found to maintain containment integrity for 50 years of storage. The validity of the acceleration test was also investigated.

Long-term storage,Metal seal,Containment integrity,Accelerated test,Stress relaxation,Creep

A multi-regression model was developed to estimate the decommissioning cost for Kori unit 1 using foreign nuclear power plant (NPP) decommissioning cost data. First, the decommissioning cost data were collected for 13 boiling water reactors and 16 pressurized water reactors and converted into the values as of November 2019. Then, for the regression model, the decommissioning cost was chosen as the dependent variable, and two variables were selected as independent variables: a contamination factor that was designed to reflect the operational characteristics of the decommissioned NPP and the decommissioning period. A statistical package in the R language was used to derive the regression model. Finally, the regression model was applied to estimate the decommissioning cost for Kori unit 1. The estimated decommissioning cost for Kori unit 1 was 663.40~928.32 million US dollars (782,812~1,095,418 million Korean won).

Regression model,Nuclear power plant,Decommissioning cost,Kori unit 1,R language

North Korea conducted the sixth underground nuclear test on September 3, 2017 at the Punggye-ri Nuclear Test Site (NTS). In contrast to the previous five nuclear tests, several induced earthquakes occurred around the NTS after the sixth nuclear test and this may have caused radioxenon leakages at the site. Considering these reported earthquakes, we performed atmospheric dispersion simulations on some radioxenon emission scenarios for this event using our Lagrangian Atmospheric Dose Assessment System (LADAS) model by employing the Unified Model (UM) based numerical weather prediction data produced by the Korea Meteorological Administration (KMA). To find out possible detection locations and times, we combined not only daily and weekly based delayed releases but also leakages after the reported earthquakes around the NTS to create emission scenarios. Our simulation results were generally in good agreement with the measured data of the Nuclear Safety and Security Commission and International Monitoring System (IMS) stations operated by the Comprehensive nuclear Test-Ban-Treaty Organization (CTBTO).

Radioxenon,Atmospheric dispersion,Induced earthquake,IMS,CTBTO

In 2017, multimodal transportation tests for evaluating road, sea, and rail transport were performed by research institutes in the US, Spain, and the Republic of Korea. In this study, acceleration and strain data determined through road and sea tests were analyzed. It was investigated whether the load generated for each transport mode was amplified or attenuated according to the load transfer path. From the results, it was confirmed that the load transfer characteristics differed according to the transportation mode and loading path. The effects of strain determined through each test on the structural integrity of the spent nuclear fuel were also investigated. It was found that the magnitude of the measured strain had a negligible effect on the structural integrity of the spent nuclear fuel, considering its fatigue strength. The results for the acceleration and strain data analyses obtained in this study will be useful for scheduled domestic transportation tests under normal transport conditions.

Normal condition of transport,Vibration and shock,Spent nuclear fuel,Structural integrity,Transportability,Durability

The Kori-Unit 1 nuclear power plant, which is scheduled to be decommissioned after permanent shutdown, is expected to generate large amounts of various types of radioactive waste during the decommissioning process. Among these, nuclear reactors and internal structures have high levels of radioactivity and the dismantled structure must have the proper size and weight on the primary side. During decommissioning, it is important to prepare an appropriate and efficient disposal method through analysis of the disposal status and the legal restrictions on wastes generated from the reactors and internal structures. Nuclear reactors and internal structures generate radioactive wastes of various levels, such as medium, very low, and clearance. A radiation evaluation indicates that wastes in the clearance level are generated in the reactor head and upper head insulation. In this study, a clearance waste safety evaluation was conducted using the RESRAD-RECYCLE code, which is a safety evaluation code, based on the activation evaluation results for the clearance level wastes. The clearance scenario for the target radioactive waste was selected and the maximum individual and collective exposure doses at the time of clearance were calculated to determine whether the clearance criteria limit prescribed by the Nuclear Safety Act was satisfied. The evaluation results indicated that the doses were significantly low, and the clearance criteria were satisfied. Based on the safety assessment results, an appropriate metal recycle and disposal method were suggested for clearance, which are the subject of the deregulation of internal structures of nuclear power plant.

Decommissioning,RV/RVI,Exposure dose,Clearance wastes,RESRAD-RECYCLE

For geological disposal of radioactive wastes, a method was proposed to evaluate the radionuclide transport in the biosphere by calculating the elapsed time of nuclide migration. The radionuclides were supposed to be introduced from a natural barrier and reached a large surface water body following a groundwater flow in a shallow subsurface. The biosphere was defined as a shallow subsurface environment that included aquifers on a host rock. Using the proposed method, a calculation algorithm was established, and a computer code that implemented the algorithm was developed. The developed code was verified by comparing the simulation results of the simple cases with the results of the analytical solution and a public program, which has been widely used to evaluate the radiation dose using the radionuclide transport near the surface. A case study was constructed using the previous research for radionuclide transport from the hypothetical geological disposal repository. In the case study, the code calculated the mass discharge rate of radionuclide to a stream in the biosphere. Because the previous research only demonstrated the transport of radionuclides from the hypothetical repository to the host rock, the developed code in the present study could help identify the total transport of radionuclide along the complete pathway.

Radionuclide transport simulation,Transport in porous media,Radionuclide transport in biosphere,Timedomain random walk (TDRW) method

The Korea Radioactive Waste Agency plans to expand the storage capacity of radioactive waste by constructing a radioactive waste inspecting building to solve the problem of the lack of inspection space and drum-handling space in the radioactive waste receipt and storage building for the first-stage disposal facility. In this study, the exposure doses of radiation workers that handle new disposal containers for decommissioning waste in the storage areas of the radioactive waste inspecting building were calculated using the Monte Carlo N-particle transport code. The annual collective dose was calculated as a total of 84.8 man-mSv for 304 new disposal containers and an estimated annual 306 working hours for the radiation work. When the 304 new disposal containers (small/medium type) were stored in the storage areas, it was found that 25 radiation workers should be involved in acceptance/disposal inspection, and the estimated exposure dose per worker was calculated as an average annual value of 3.39 mSv. When the radiation workers handle the small containers in high-radiation dose areas, the small containers should be shielded further by increasing the concrete liner thickness to improve the work efficiency and radiation safety of the radiation workers. The results of this study will be useful in establishing the optimal radiation working conditions for radiation workers using the source term and characteristics of decommissioning waste based on actual measurements.

Low and intermediate-level radioactive waste,Decommissioning waste,Package/disposal container,Radioactive waste inspecting building,Radiation worker,Exposure dos

The high-flux advanced neutron application reactor (HANARO) is a research reactor with thermal power of 30 MW applied in various research and development using neutrons generated from uranium fission chain reaction. A degasifier tank is installed in the ancillary facility of HANARO. This facility generates gas pollutants produced owing to internal environmental factors. The degasifier tank is designed to maintain the gas contaminants below acceptable levels and is monitored using an analyzer in the gas sampling panel. If condensate water is generated and flows into the analyzer of the gas sampling panel, corrosion occurs inside the analyzer';s measurement chamber, which causes failure. Condensate water is generated because of the temperature difference between the degasifier tank and analyzer when the gas flows into the analyzer. A heating system is installed between the degasifier tank and gas sampling panel to suppress condensate water generation and effectively remove the condensate water inside the system. In this study, we investigated the efficiency of the heating system. In addition, the variations in the pipe temperature and the amount of average condensate water were modeled using a wall condensation model based on the changes in the fluid inlet temperature, outside air temperature, and heating cable-setting temperature.

HANARO (high-flux advanced neutron application reactor),Condensate water,CFD (computational fluid dynamics),Wall condensation model,Heating system

Two commonly used ceramics in molten salt research are alumina and mullite. The two ceramics were exposed to a combination of rare earth chlorides (YCl₃, SmCl₃, NdCl₃, PrCl₃, and CeCl₃; each rare earth chloride of 1.8 weight percent) in LiCl-KCl at 773 K for approximately 13 days. Scanning electron microscopy with wave dispersion spectra was utilized to investigate a formation layer or deposition of rare earths onto the ceramic. Only the major constituents of the ceramics (Al, Si, and O₂) were observed during the wave dispersion spectra. X-ray fluorescence was used as well to determine concentration changes in the molten salt as a function of ceramic exposure time. This study shows no evidence of ionic exchange or layer formation between the ceramics and molten chloride salt mixture. There are signs of surface tension effects of molten salt moving out of the tantalum crucible into secondary containment.

Molten salt,Ceramic,Ionic exchange,Rare earth chloride,Molten salt surface tension

Site characterization for decommissioning Kori Unit 1 is ongoing in South Korea after 40 years of successful operation. Kori Unit 1’s containment building is assumed to be mostly radioactively contaminated, and therefore radiation exposure management and detailed contamination investigation are required for decommissioning and dismantling it safely. In this study, site-specific Derived Concentration Guideline Levels (DCGLs) were derived using the residual radioactivity risk evaluation tool, RESRAD-BUILD code. A conceptual model of containment building for Kori Unit 1 was set up and limited occupational worker building inspection scenario was applied. Depending on the source location, the maximum contribution source and exposure pathway of each radionuclide were analyzed. The contribution of radionuclides to dose and exposure pathways, by source location, is expected to serve as basic data in the assessment criteria of survey areas and classification of impact areas during further decommissioning and decontamination of sites.

MARSSIM,Containment building,DCGL,RESRAD-BUILD,Pathway analysis

The corrosion behavior of the Inconel X-750 alloy was investigated for its potential application under a Cl₂-O₂ mixed gas flow in an Ar atmosphere. The corrosion rate was found to be negligible at temperatures up to 400℃ under a flow rate of 30 mL·min^-1 Cl₂ + 170 mL·min^-1 Ar, whereas an exponential increase was observed in the corrosion rate at temperatures greater than 500℃. The suppression of the corrosion reaction due to the presence of O₂ was verified experimentally at flow rates of 30 mL·min^-1 Cl₂ (4.96 g·m^-2·h^-1), 20 mL·min^-1 Cl₂ +10 mL·min^-1 O₂ (2.02 g·m^-2·h^-1), and 10 mL·min^-1 Cl₂ + 20 mL·min^-1 O₂ (1.34 g·m^-2·h^-1) under a constant Ar flow rate of 170 mL·min^-1 at 600℃ for 8 h. The surface morphology analysis results revealed that porous surfaces with tunnel-type holes were produced under the Cl₂-O₂ mixed-gas condition. Furthermore, the effects of the Cl₂ flow rate on the corrosion rate were investigated, indicating that its impact was negligible within the range of 5–30 mL·min^-1 Cl₂ at 600℃.

Inconel X-750,Chlorine,Oxygen-chlorine mixed gas,Oxide reduction,Pyroprocessing

Anionic radionuclides pose one of the highest risks to the long-term safety assessments of disposal repositories. Therefore, techniques to immobilize and separate such anionic radionuclides are of crucial importance from the viewpoints of safety and waste volume reduction. The main objective of this study is to design a separator with minimum pressure disturbance, based on the concept of a conventional cyclone separator. We hypothesize that the anionic radionuclides can be immobilized onto a nanomaterial-based substrate and that the particles generated in the process can flow via water. These particles are denser than water; hence, they can be trapped within the cyclone-type separator because of its design. We conducted particle tracking analysis using computational fluid dynamics (CFD) for the conventional cyclone separator and studied the effects due to the morphology of the separator. The proposed sandglass-like design of the separator shows promising results (i.e., only one out of 10,000 particles escaped to the outlet from the separation zone). To validate the design, we manufactured a laboratory-scale prototype separator and tested it for iron particles; the efficiency was ca. 99%. Furthermore, using an additional magnetic effect with the separator, we could effectively separate particles with ~100% efficiency. The proposed sandglass-like separator can thus be used for effective separation and recovery of immobilized anionic radionuclides.

Anionic radionuclide,Separation,Sandglass-like separator,Particle tracking

This study proposes a method of separating uranium (U) and minor actinides from rare earth (RE) elements in the LiCl-KCl salt system. Several RE metals were used to reduce UCl₃ and MgCl₂ from the eutectic LiCl-KCl salt systems. Five experiments were performed on drawdown U and plutonium (Pu) surrogate elements from RECl₃-enriched LiCl-KCl salt systems at 773 K. Via the introduction of RE metals into the salt system, it was observed that the UCl₃ concentration can be lowered below 100 ppm. In addition, UCl₃ was reduced into a powdery form that easily settled at the bottom and was successfully collected by a salt distillation operation. When the RE metals come into contact with a metallic structure, a galvanic interaction occurs dominantly, seemingly accelerating the U recovery reaction. These results elucidate the development of an effective and simple process that selectively removes actinides from electrorefining salt, thus contributing to the minimization of the influx of actinides into the nuclear fuel waste stream.

Actinide drawdown,Pyroprocess,Uranium,Electrorefiner,LiCl-KCl

The corrosion behavior of Hastelloy C-276 was investigated to identify its applicability for carbon-anode-based oxide reduction (OR), in which Cl₂ and O₂ are simultaneously evolved at the anode. Under a 30 mL·min^-1 Cl₂ + 170 mL·min-1 Ar flow, the corrosion rate was less than 1 g·m^-2·h^-1 up to 500℃, whereas the rate increased exponentially from 500 to 700℃. The effects of the Cl₂-O₂ composition on the corrosion rate at flow rates of 30 mL·min^-1 Cl₂, 20 mL·min^-1 Cl₂ + 10 mL·min^-1 O₂, and 10 mL·min^-1 Cl₂ + 20 mL·min^-1 O₂ with a constant 170 mL·min^{-1 Ar flow rate at 600℃ was analyzed. Based on the data from an 8 h reaction, the fastest corrosion rate was observed for the 20 mL·min-1 Cl₂ + 10 mL·min-1 O₂ case, followed by 30 mL·min-1 Cl₂ and 10 mL·min-1 Cl₂ + 20 mL·min-1 O₂. The effects of the chlorine flow rate on the corrosion rate were negligible within the 5–30 mL·min-1 range. A surface morphology analysis revealed the formation of vertical scratches in specimens that reacted under the Cl₂-O₂ mixed gas condition.}

Hastelloy C-276,Corrosion,Oxide reduction,Chlorine-oxygen mixed gas,Chlorine

Compartment fire tests were performed using kerosene and Jet A-1 as fire sources to evaluate the relationship between flame temperature and opening size. The tests were performed for a fire caused by the release of kerosene owing to vehicle impact, and for a fire caused by the release of Jet-A-1 owing to airplane collision. The compartment fire tests were performed using a 1/3-scale model of a metal storage cask when the flame temperature was deemed to be the highest. We found the combustion time of Jet-A-1 to be shorter than that of kerosene, and consequently, the flame temperature of Jet-A-1 was measured to be higher than that of kerosene. When the opening was installed on the compartment roof, even though the area of the opening was small, the ventilation factor was large, resulting in a high flame temperature and long combustion. Therefore, the position of the opening is a crucial factor that affects the flame temperature. When the metal storage cask was stored in the compartment, the flame temperature decreased proportionally with the energy that the metal storage cask received from the flame.

Compartment fire,Flame temperature,Opening,Metal storage cask

Thermal decomposition of the uranyl phosphate mineral phase meta-ankoleite (KUO₂PO₄·3H₂O) has been considered in relation to high temperature thermal sintering for the immobilisation of a uranyl phosphate containing waste. Meta-ankoleite thermal decomposition was studied across the temperature range 25 – 1200℃ under an inert N2 atmosphere at 1 atm. It is shown that the meta-ankoleite mineral phase undergoes a double de-hydration event at 56.90 and 125.85℃. Subsequently, synthetically produced pure meta-ankoleite remains stable until at least 1150℃ exhibiting no apparent phase changes. In contrast, when present in a mixed waste the meta-ankoleite phase is not identifiable after thermal treatment indicating incorporation within the bulk waste either as an amorphous phase and/or as uranium oxide. Visual inspection of the waste post thermal treatment showed evidence of self-sintering owing to the presence of glass former materials, namely, silica (SiO₂) and antimony(V) oxide (Sb₂O₅). Therefore, incorporation of the uranium phase into the waste as part of waste sentencing and immobilisation via high temperature sintering for the purpose of long-term disposal is deemed feasible.

Uranium,Uranyl phosphate,Meta-ankoleite,Thermogravimetry,Radioactive waste,SOHIO process

Strontium is known as a salt-soluble element during the electrolytic oxide reduction (EOR) process. The chemical behavior of SrO during EOR was investigated via thermodynamic calculations to provide quantitative data on the chemical status of Sr. To achieve this, thermodynamic calculations were conducted using HSC chemistry software for various EOR conditions. It was revealed that SrO reacts with LiCl salt to produce SrCl₂, even in the presence of Li₂O, and that the ratio of SrCl₂ depends on the initial concentration of Li₂O dissolved in LiCl. It was found that SrO reacts with Li to produce Sr during EOR and that the reduced Sr reacts with LiCl salt to produce SrCl₂. As a result, the proportions of metallic forms were lower in Sr than in La and Nd under various EOR conditions. The thermodynamic calculations indicated that the three chemical forms of SrO, SrCl₂, and Sr co-exist in the EOR system under an equilibrium with Li, Li₂O, and LiCl.

Strontium oxide,Thermodynamic calculations,Electrolytic oxide reduction,Pyroprocessing,HSC chemistry

This study provides an assessment on a proposed method for separation of cesium, strontium, and barium using electrochemical reduction at a liquid bismuth cathode in LiCl-KCl eutectic salt, investigated via cyclic voltammetry (CV), electrochemical impedance spectroscopy (EIS), and scanning electron microscopy with energy dispersive X-ray spectrometry (SEM-EDS). CV studies were performed at temperatures of 723-823 K and concentrations of the target species up to 4.0wt%. Redox reactions occurring during potential sweeps were observed. Concentration of BaCl₂ in the salt did not seem to influence the diffusivity in the studied concentration range up to 4.0wt%. The presence of strontium in the system affected the redox reaction of lithium; however, there were no distinguishable redox peaks that could be measured. Impedance spectra obtained from EIS methods were used to calculate the exchange current densities of the electroactive active redox couple at the bismuth cathode. Results show the rate-controlling step in deposition to be the mass transport of Cs⁺ions from the bulk salt to the cathode surface layer. Results from SEM-EDS suggest that Cs-Bi and Sr-Bi intermetallics from LiCl-KCl salt are not thermodynamically favorable.

Liquid bismuth cathode,Electrochemical impedance spectxoscopy,Electrochemistry,Cesium,Strontium,Barium

Numerical model was developed that simulates radionuclide (³H and ¹⁴C) transport modeling at the 2^nd phase facility at the Wolsong LILW Disposal Center. Four scenarios were simulated with different assumptions about the integrity of the compo-nents of the barrier system. For the design case, the multi-barrier system was shown to be effective in diverting infiltration water around the vaults containing radioactive waste. Nevertheless, the volatile radionuclide ¹⁴C migrates outside the con-tainment system and through the unsaturated zone, driven by gas diffusion. ³H is largely contained within the vaults where it decays, with small amounts being flushed out in the liquid state. Various scenarios were examined in which the integrity of the cover barrier system or that of the concrete were compromised. In the absence of any engineered barriers, ³H is washed out to the water table within the first 20 years. The release of ¹⁴C by gas diffusion is suppressed if percolation fluxes through the facility are high after a cover failure. However, the high fluxes lead to advective transport of 14C dissolved in the liquid state. The concrete container is an effective barrier, with approximately the same effectiveness as the cover.

Low and intermediate level radioactive waste,Radionuclides migration,Second phase facility,Long-term safety

In this study, extraction of uranium(VI) from an aqueous nitric acid solution was investigated using tri-n-butyl phosphate (TBP) as an extractant in an ionic liquid, 1-alkyl-3-methylimidazolium bis (trifluoromethylsulfonyl)imide ([C_nmim][Tf₂N]). The distribution ratio of U(VI) in 1.1 M TBP/[C_nmim][Tf₂N] was significantly high when the concentration of nitric acid was low. The value of the distribution ratio decreased as the concentration of the nitric acid increased at lower acidities, and then increased with a nitric acid concentration of up to 8 M. This can be attributed to the different extraction mechanisms of U(VI) based on nitric acid concentrations. Thus, a cation exchange at low acidity levels and an ion-pair extraction at high acidity levels were suggested as the extraction mechanism of U(VI) in the TBP/[C_nmim][Tf₂N] system.

Extraction,Uranium,TBP,Ionic liquid

Two waste forms, namely cement and geopolymer, were investigated and tested in this study to solidify the corrosive sludge generated from the surface and precipitates of the tubes of steam generators in nuclear power plants. The compressive strength of the cement waste form cured for 28 days was inversely proportional to waste loading (24.4 MPa for 0wt% to 2.7 MPa for 60wt%). The corrosive sludge absorbed the free water in the hydration reaction to decrease the cementation reaction. When the corrosive sludge waste loading increased to 60wt%, the cement waste form showed decreased compressive strength (2.7 MPa), which did not satisfy the acceptance criteria of the repository (3.45 MPa). Meanwhile, the compressive strength of the geopolymer waste form cured for 7 days was proportional to waste loading (23.6 MPa for 0wt% to 31.9 MPa for 40wt%). The corrosive sludge absorbed the free water in the geopolymer when the water content decreased, such that a compact geopolymer structure could be obtained. Consequently, the geopolymer waste forms generally showed higher compressive strengths than cement waste forms.

Cement,Geopolymer,Waste form,Waste loading,Compressive strength,Sludge waste

After nuclear power plants are permanently shut down and decommissioned, the remaining irradiated metal components such as stainless steel, carbon steel, and Inconel can be used as neutron absorber. This study investigates the possibility of reusing these metal components as neutron absorber materials, that is burnable poison. The absorption cross section of the irradiated metals did not lose their chemical properties and performance even if they were irradiated over 40-50 years in the NPPs. To examine the absorption capability of the waste metals, the lattice calculations of WH 17×17 fuel assembly were analyzed. From the results, Inconel-718 significantly hold-down fuel assembly excess reactivity compared to stainless steel 304 and carbon steel because Inconel-718 contains a small amount of boron nuclide. From the results, a 20wt% impurity of boron in irradiated Inconel-718 enhances the excess reactivity suppression. The application of irradiated Inconel-718 as a burnable absorber for SMR core was investigated. The irradiated Inconel-718 impurity with 20wt% of boron content can maintain and suppress the whole core reactivity. We emphasize that the irradiated metal components can be used as burnable absorber materials to control the reactivity of commercial reactor power and small modular reactors.

Reusable metal components,Scrap metal,Irradiated Inconel-718,Burnable absorber,Monte Carlo simulation

In the geochemical field, the chemical speciation of hexavalent uranium (U(VI)) has been widely investigated by performing measurements to determine its luminescence properties, namely the excitation, emission, and lifetime. Of these properties, the excitation has been relatively overlooked in most time-resolved laser fluorescence spectroscopy (TRLFS) studies. In this study, TRLFS and continuous-wave excitation–emission matrix spectroscopy are adopted to characterize the excitation properties of U(VI) surface species that interact with amorphous silica. The luminescence spectra of U(VI) measured from a silica suspension and silica sediment showed very similar spectral shapes with similar lifetime values. In contrast, the excitation spectra of U(VI) measured from these samples were significantly different. The results show that distinctive excitation maxima appeared at approximately 220 and 280 nm for the silica suspension and silica sediment, respectively.

Time-resolved laser fluorescence spectroscopy (TRLFS),Excitation–emission matrix (EEM) spectroscopy,Adsorption,Hexavalent uranium,Amorphous silica

An important property of glass and ceramic solid waste forms is processability. Tellurite materials with low melting temperatures and high halite solubilities have potential as solid waste forms. Crystalline TiTe₃O₈ was synthesized through a solid-state reaction between stoichiometric amounts of TiO₂ and TeO₂ powder. The resultant TiTe₃O₈ crystal had a three-dimensional (3D) structure consisting of TiO₆ octahedra and asymmetric TeO₄ seesaw moiety groups. The melting temperature of the TiTe₃O₈ powder was 820℃, and the constituent TeO₂ began to evaporate selectively from TiTe₃O₈ above around 840℃. The leaching rate, as determined using the modified American Society of Testing and Materials static leach test method, of Ti in the TiTe₃O₈ crystal was less than the order of 10^-4 g·m^-2·d^-1 at 90℃ for durations of 14 d over a pH range of 2-12. The chemical durability of the TiTe₃O₈ crystal, even under highly acidic and alkaline conditions, was comparable to that of other well-known Ti-based solid waste forms.

Solid waste form,Solid-state reaction,TiTe_{3<／sub>O8<／sub>,Leaching rate,Chemical durability}

The radionuclide inventory in radioactive waste from nuclear power plants should be determined to secure the safety of final repositories. As an alternative to time-consuming, labor-intensive, and destructive radiochemical analysis, the indirect scaling factor (SF) method has been used to determine the concentrations of difficult-to-measure radionuclides. Despite its long history, the original SF methodology remains almost unchanged and now needs to be improved for advanced SF implementation. Intense public attention and interest have been strongly directed to the reliability of the procedures and data regarding repository safety since the first operation of the low- and intermediate-level radioactive waste disposal facility in Gyeongju, Korea. In this review, statistical methodologies for SF implementation are described and evaluated to achieve reasonable and advanced decision-making. The first part of this review begins with an overview of the current status of the scaling factor method and global experiences, including some specific statistical issues associated with SF implementation. In addition, this review aims to extend the applicability of SF to the characterization of large quantities of waste from the decommissioning of nuclear facilities.

Scaling factor,Radioactive waste,Waste characterization,Decision criteria,Statistical methodology

Currently, radioactive waste for disposal has been restricted to low and intermediate level radioactive waste generated during operation of nuclear power plants, and these radioactive wastes were managed and disposed of the 200 L and 320 L of steel drums. However, it is expected that it will be difficult to manage a large amount of decommissioning waste of the Kori unit 1 with the existing drums and transportation containers. Accordingly, the KORAD is currently developing various and large-sized containers for packaging, transportation, and disposal of decommissioning waste. In this study, the radiation exposure doses of workers and the public were evaluated using RADTRAN computational analysis code in case of the domestic on-road transportation of new package and transportation containers under development. The results were compared with the domestic annual dose limit. In addition, the sensitivity of the expected exposure dose according to the change in the leakage rate of radionuclides in the waste packaging was evaluated. As a result of the evaluation, it was confirmed that the exposure dose under normal and accident condition was less than the domestic annual exposure dose limit. However, in the case of a number of loading and unloading operations, working systems should be prepared to reduce the exposure of workers.

Decommissioning waste,Transportation container,RADTRAN,On-road transportation,Annual dose limit

Deionized water, methanol, and ethanol were investigated for their effectiveness at dissolving LiCl-KCl-UCl₃ at 25, 35, and 50℃ using inductively coupled plasma mass spectrometry (ICP-MS) to study the concentration evolution of uranium and mass ratio evolutions of lithium and potassium in these solvents. A visualization experiment of the dissolution of the ternary salt in solvents was performed at 25℃ for 2 min to gain further understanding of the reactions. Aforementioned solvents were evaluated for their performance on removing the adhered ternary salt from uranium dendrites that were electrochemically separated in a molten LiCl-KCl-UCl₃ electrolyte (500℃) using scanning electron microscopy with energy dispersive spectroscopy (SEM-EDS). Findings indicate that deionized water is best suited for dissolving the ternary salt and removing adhered salt from electrodeposits. The maximum uranium concentrations detected in deionized water, methanol, and ethanol for the different temperature conditions were 8.33, 5.67, 2.79 μg·L^-1 for 25℃, 10.62, 5.73, 2.50 μg·L^-1 for 35℃, and 11.55, 6.75, and 4.73 μg·L^-1 for 50℃. ICP-MS analysis indicates that ethanol did not take up any KCl during dissolutions investigated. SEM-EDS analysis of ethanol washed uranium dendrites confirmed that KCl was still adhered to the surface. Saturation criteria is also proposed and utilized to approximate the state of saturation of the solvents used in the dissolution trials.

Chloride Salt Dissolution,Electrodeposited Uranium,SEM-EDS,ICP-MS

Based on the current high-level radioactive waste management basic plan and the analysis results of spent nuclear fuel characteristics, such as dimensions and decay heat, an improved geological disposal concept for spent nuclear fuel from domestic nuclear power plants was proposed in this study. To this end, disposal container concepts for spent nuclear fuel from two types of reactors, pressurized water reactor (PWR) and Canada deuterium uranium (CANDU), considering the dimensions and interim storage method, were derived. In addition, considering the cooling time of the spent nuclear fuel at the time of disposal, according to the current basic plan-based scenarios, the amount of decay heat capacity for a disposal container was determined. Furthermore, improved disposal concepts for each disposal container were proposed, and analyses were conducted to determine whether the design requirements for the temperature limit were satisfied. Then, the disposal efficiencies of these disposal concepts were compared with those of the existing disposal concepts. The results indicated that the disposal area was reduced by approximately 20%, and the disposal density was increased by more than 20%.

Spent nuclear fuel,Disposal container,Deep geological disposal system,Disposal scenario,Disposal efficiency

With respect to spent nuclear fuels, disposal containers and bentonite buffer blocks in deep geological disposal systems are the primary engineered barrier elements that are required to isolate radioactive toxicity for a long period of time and delay the leakage of radio nuclides such that they do not affect human and natural environments. Therefore, the thermal stability of the bentonite buffer and structural integrity of the disposal container are essential factors for maintaining the safety of a deep geological disposal system. The most important requirement in the design of such a system involves ensuring that the temperature of the buffer does not exceed 100℃ because of the decay heat emitted from high-level wastes loaded in the disposal container. In addition, the disposal containers should maintain structural integrity under loads, such as hydraulic pressure, at an underground depth of 500 m and swelling pressure of the bentonite buffer. In this study, we analyzed the thermal stability and structural integrity in a deep geological disposal environment of the improved deep geological disposal systems for domestic light-water and heavy-water reactor types of spent nuclear fuels, which were considered to be subject to direct disposal. The results of the thermal stability and structural integrity assessments indicated that the improved disposal systems for each type of spent nuclear fuel satisfied the temperature limit requirement (< 100℃) of the disposal system, and the disposal containers were observed to maintain their integrity with a safety ratio of 2.0 or higher in the environment of deep disposal.

Spent nuclear fuel,Deep geological disposal,Design requirements,Bentonite block,Disposal container,Thermal stability,Structural integrity

In this paper, an overview of the scoping calculation results is provided with respect to criticality and radiation shielding of two KBS-3V type PWR SNF disposal systems and one NWMO-type CANDU SNF disposal system of the improved KAERI reference disposal system for SNFs (KRS⁺). The results confirmed that the calculated effective multiplication factors (k_eff) of each disposal system comply with the design criteria (< 0.95). Based on a sensitivity study, the bounding conditions for criticality assumed a flooded container, actinide-only fuel composition, and a decay time of tens of thousands of years. The necessity of mixed loading for some PWR SNFs with high enrichment and low discharge burnup was identified from the evaluated preliminary possible loading area. Furthermore, the absorbed dose rate in the bentonite region was confirmed to be considerably lower than the design criterion (< 1 Gy‧hr^-1). Entire PWR SNFs with various enrichment and discharge burnup can be deposited in the KRS⁺ system without any shielding issues. The container thickness applied to the current KRS⁺ design was clarified as sufficient considering the minimum thickness of the container to satisfy the shielding criterion. In conclusion, the current KRS⁺ design is suitable in terms of nuclear criticality and radiation shielding.

Criticality,Radiation shielding,Spent nuclear fuel,Disposal system,Geological repository

In this paper, an approach developed by the Finnish nuclear waste management organization, Posiva, for the construction license of a geological repository was reviewed. Furthermore, a computer program based on the approach was developed. By using the computer program, the lifetime of a copper disposal canister, which was a key engineered barrier of the geological repository, was predicted under the KAERI Underground Research Tunnel (KURT) geologic conditions. The computer program was developed considering the mass transport of corroding agents, such as oxygen and sulfide, through the buffer and backfill. Shortly after the closure of the repository, the corrosion depths of a copper canister due to oxygen in the pores of the buffer and backfill were calculated. Additionally, the long-term corrosion of a copper canister due to sulfide was analyzed in two cases: intact buffer and eroded buffer. Under various conditions of the engineered barrier, the corrosion lifetimes of the copper canister due to sulfide significantly exceeded one million years. Finally, this study shows that it is necessary to carefully characterize the transmissivity of rock and sulfide concentration during site characterization to accurately predict the canister lifetime.

Spent nuclear fuel,Disposal,Copper canister,Corrosion,Buffer

The operational safety assessment is an important part of a safety case for the deep geological repository of spent fuels. It consists of different stages such as the identification of initiating events, event tree analysis, fault tree analysis, and evaluation of exposure doses to the public and radiation workers. This study develops a probabilistic safety assessment method for the operational safety assessment and establishes an assessment framework. For the event and fault tree analyses, we propose the advanced information management system for probabilistic safety assessment (AIMS-PSA Manager). In addition, we propose the Radiological Safety Analysis Computer (RSAC) program to evaluate exposure doses to the public and radiation workers. Furthermore, we check the applicability of the assessment framework with respect to drop accidents of a spent fuel assembly arising out of crane failure, at the surface facility of the KRS⁺ (KAERI Reference disposal System for SNFs). The methods and tools established through this study can be used for the development of a safety case for the KRS⁺ system as well as for the design modification and the operational safety assessment of the KRS⁺ system.

Spent fuel,Deep geological repository,Operational safety assessment,Probabilistic safety assessment,Exposure dose

The Korea Atomic Energy Research Institute (KAERI) has developed geological repository systems for the disposal of high-level wastes and spent nuclear fuels (SNFs) in South Korea. The purpose of the most recently developed system, the improved KAERI Reference Disposal System Plus (KRS⁺), is to dispose of all SNFs in Korea with improved disposal area efficiency. In this paper, a system-level safety assessment model for the KRS⁺ is presented with long-term assessment results. A system-level model is used to evaluate the overall performance of the disposal system rather than simulating a single component. Because a repository site in Korea has yet to be selected, a conceptual model is used to describe the proposed disposal system. Some uncertain parameters are incorporated into the model for the future site selection process. These parameters include options for a fractured pathway in a geosphere, parameters for radionuclide migration, and repository design dimensions. Two types of SNF, PULS7 from a pressurized water reactor and Canada Deuterium Uranium from a heavy water reactor, were selected as a reference inventory considering the future cumulative stock of SNFs in Korea. The highest peak radiological dose to a representative public was estimated to be 8.19×10^-4 mSv‧yr^-1, primarily from ¹²⁹I. The proposed KRS⁺ design is expected to have a high safety margin that is on the order of two times lower than the dose limit criterion of 0.1 mSv‧yr^-1.

Spent nuclear fuel,Disposal system,Geological repository,Safety assessment,KAERI reference system