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.
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.
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.
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 Ar-41, Kr-85, and Xe-133 were occasionally emitted. Among gaseous effluents, H-3,C- 14, Ar-41, Xe-133, Xe-131(m), and Kr-85 showed high frequencies. For liquid effluents, 3H, 60Co, 58Co, and 95Nb were the most frequently released. In terms of dose contribution, C-14, H-3,Xe- 133, Ar-41, and Co-60 were dominant among gaseous effluents, while C-14, H-3, Nb-95, Co-58, and Co-60 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.
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.
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.
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 & centerdot;tU-1) were sintered at 1,550 degrees C or 1,700 degrees C, then ground and tested using the Product Consistency Test A at 90 degrees 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 1,700 degrees C, with each element's behavior reflecting its microstructural association. epsilon-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.
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 & centerdot;g-1) was lower than that of pristine NiAl LDH powder (0.6750 mmol & centerdot;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.
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 & centerdot;g-1 for all three isotopes, whereas domestic assessments yielded 1 Bq & centerdot;g-1 for 234U and 238U and 0.1 Bq & centerdot;g-1 for 235U. Because uranium isotopes typically occur 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 & centerdot;g-1 assumption and the case applying 0.1 Bq & centerdot;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 & centerdot;g-1 for 234U, 235U, and 238U is recommended as reasonable and technically justified.
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.
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.
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.
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.
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 x 10-13 ms-1 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.
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.
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 30wt% with a compressive strength of 16.607 MPa. Uranium precipitates containing about 40wt% diatomite were immobilized in a cement matrix at a maximum loading of 20wt%, 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 30wt% 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.
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.
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 spentfuel 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 similar to 9% increase in impact load and for an similar to 8% reduction in rack-cell thickness, indicating potential buckling only beyond those thresholds. These findings support safe SFP operation with high-density configurations.
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.
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.
