Journal of Nuclear Fuel Cycle and Waste Technology

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