Journal of Nuclear Fuel Cycle and Waste Technology

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

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

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

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

Radioactive iodine isotopes (¹²⁹I and ¹³¹I) from spent nuclear fuel pose significant environmental risks due to high radioactivity and mobility in aqueous systems. This study embedded NiAl Layered Double Hydroxide (LDH) within sodium alginate and poly vinyl alcohol matrices by crosslinking with CaCl₂ to fabricate bead-type sorbents for I– removal. XRD and FT-IR analyses confirmed that the crystallinity of NiAl LDH was retained within the beads, indicating structural stability. However, the sorption capacity of NiAl LDH beads (0.2151–0.2489 mmol·g^–1) was lower than that of pristine NiAl LDH powder (0.6750 mmol·g^–1), primarily due to partial anion-exchange of interlayer NO₃– by Cl– during bead formation, as Cl– has a higher affinity than NO₃–. Despite this, effective I– sorption occurred. Zeta potential measurements revealed an increase in surface potential after I– sorption, which contradicted the typical behavior of electrostatic attraction. This suggests that structural rearrangement of the bead, driven by Na⁺–Ca²⁺ exchange under NaI used for I– solution, may have led to increased LDH surface exposure. This exposure enabled I– sorption via anion-exchange, allowing partial substitution of interlayer anions. These findings can offer insights for the design of bead-type sorbents optimized for radioactive iodine removal.

Iodide, LDH, Bead, Anion-exchange, Specific ion effect

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

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 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

This technical paper documents an assessment of the structural integrity of an individual spent fuel rod under a hypothetical drop accident occurring during fuel-bundle handling in the spent fuel pool (SFP) at domestic pressurized heavy-water reactor (PHWR) nuclear power plants. The analysis is grounded in the operating procedures for managing and transferring spent fuel (SF) in the SFP. Although an anti-drop metallic screen is installed on actual pool floors, a rigid floor was assumed for conservatism. The evaluation encompasses determining the underwater drop velocity with due consideration of hydrodynamic damping in the pool environment, estimating the resulting impact load, and calculating stresses and strains in the Zircaloy-4 cladding(sheath), followed by comparison with acceptance criteria for preventing failure and leakage and the associated design margins. Conservative assumptions were applied for impact location, drop height, and material properties; the methodology and input assumptions are presented, and representative case calculations are provided to confirm whether structural integrity is maintained.

PHWR, Spent fuel, SFP, Fuel bundle, Fuel rod drop, Zircalloy-4 cladding

Clearance serves as a regulatory mechanism to reduce unnecessary restrictions, lower economic costs, promote recycling of metallic resources, minimize decommissioning expenses, and improve efficiency of radioactive waste disposal facilities. In South Korea, clearance is determined by compliance with clearance dose or clearance level (CL). However, CLs for uranium isotopes (²³⁴U, ²³⁵U, ²³⁸U) have not yet been established, requiring case-by-case dose assessments and creating administrative challenges. In this study, CLs for uranium isotopes were derived using the methodology of IAEA SRS No. 44 and scenarios tailored to domestic conditions. The IAEA based evaluation produced CLs of 1 Bq·g⁻¹ for all three isotopes, whereas domestic assessments yielded 1 Bq·g−1 for ²³⁴U and ²³⁸U and 0.1 Bq·g⁻¹ for ²³⁵U. Because uranium isotopes typicallyoccur as mixtures, the derived values were further applied to depleted uranium (DU), natural uranium (NU), and low enriched uranium (LEU). A comparison between a uniform 1 Bq·g⁻¹ assumption and the case applying 0.1 Bq·g⁻¹ to ²³⁵U revealed increases in summed fractions of 10%, 20%, and 26% for DU, NU, and LEU, respectively. Considering conservative assumptions in both scenarios and input parameters, a uniform CL of 1 Bq·g⁻¹ for ²³⁴U, ²³⁵U, and ²³⁸U is recommended as reasonable and technically justified.

Clearance level, Dose assessment, Uranium isotopes, NORM, ²³⁴U, ²³⁵U, ²³⁸U

A roller burnishing process was employed to control the surface roughness and residual stress, offering a practical solution for mitigating chloride-induced stress corrosion cracking (CISCC) in the aging management of austenitic stainless-steel canisters. The surface roughness and residual stress before and after roller burnishing were evaluated using a flat welded plate and a scaled canister model. The surface roughness improved by 74.5% in the former model and 60% in the scaled canister model following roller burnishing. The flat welded plate exhibited compressive residual stress at the yield strength level of stainless steel after treatment. Similarly, the scaled canister model displayed compressive residual stress comparable to the yield strength of stainless steel. Thus, the roller burnishing process enhanced surface roughness and induced compressive residual stress, effectively eliminating tensile residual stress in the weld zone and heat-affected zone. This method is expected to reduce the number of crack initiation sites and prevent crack initiation from pits caused by localized corrosion, thereby improving the CISCC resistance of austenitic stainless steel.

Nuclear spent fuel, Canister, Austenitic stainless steel, Chloride-induced stress corrosion cracking, Roller burnishing process, Residual stress

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, SSR-06 (Rev.1)

This Letter announces the official correction of five potential errors in NRCDose3 Version 1.1.4, previously reported by Chang and Cheong (Journal of Nuclear Fuel Cycle and Waste Technology, 2025). Occurring in the GASPAR and LADTAP modules when applying dose factors from International Commission on Radiological Protection Publication 72, these errors have been resolved in the newly released Version 1.1.5 through code modification. Verification using the KHU Code confirmed that the updated version produces consistent results with relative errors within approximately 1%. Therefore, users are strongly advised to adopt Version 1.1.5 for accurate and reliable dose assessment. Users who continue to operate Version 1.1.4 should continue applying the correction methods provided in the previous paper to ensure analytical accuracy.

NRCDose3, Potential error, Verification, Radiation dose assessment, Radioactive effluent