Search Results (14)

The storage and monitoring of nuclear fuel, whether spent or fresh, are key components of the nuclear energy life cycle, with significant implications for safety and sustainability. With the global focus on carbon neutrality, interest in advanced management solutions is rising. This paper provides a comprehensive analysis of modern technologies for the design, storage, and monitoring of nuclear fuel, highlighting current trends and future challenges. The study encompasses both spent and fresh nuclear fuel, with a focus on radiological safety, structural integrity, and digital monitoring. Data were organized into the following categories: storage types (wet/dry), monitored parameters, surveillance technologies (sensors, AI, IoT, and Digital Twin), simulation models, and emerging directions. A comparison between fresh and spent fuel shows a clear shift toward intelligent systems using non-invasive sensors, deep-learning algorithms, and decentralized architectures (e.g., blockchain-IoT). Despite progress, challenges remain, such as limited interoperability across system generations and insufficient experimental validation. This paper provides a solid foundation for researchers, suggesting future directions that include the full integration of AI in monitoring, broader numerical simulations for reliability, and the standardization of digital interfaces. These measures could significantly enhance the safety and efficiency of nuclear fuel storage systems. Full article

Spent nuclear fuel (SNF) from the United States’ nuclear power plants has been placed in dry cask storage systems since the 1980s. Due to the lack of a clear path for permanent geological repository for SNF, consolidated and long-term storage solutions that use durable concrete and avoid current aging and licensing challenges are becoming indispensable. Ultra-high-performance concrete (UHPC) is a rapidly growing advanced concrete solution with superior mechanical and durability properties that can help realize future resilient nuclear storage facilities. Thus, the overall goal of this review study is to demonstrate the viability of UHPC as a long-term solution for future nuclear storage facilities. The paper first identifies all possible non-nuclear (environmental) and nuclear (thermal and radiation-induced) degradation mechanisms in concrete overpacks and storage modules with critical assessment and projections on UHPC performance in comparison to current conventional concrete solutions. Next, since concrete serves as a shielding material in nuclear settings, the preliminary attenuation properties of UHPC from emerging studies are synthesized along with the possible mix modifications to improve its attenuation performance. The paper identifies the major knowledge gaps to inform future research and development, aimed at rethinking the design of SNF dry storage facilities by incorporating UHPC. Full article

Dry systems are being employed worldwide as interim storage for Spent Nuclear Fuel (SNF). Despite not being designed as permanent repositories, the safe storage of SNF must still be ensured. In this framework, few experimental campaigns have been conducted in the past. However, their limited number has led to the necessity to exploit numerical simulations for the thermal characterization of the system. Since the majority of the degradation mechanisms are temperature-dependent, conducting a thermal analysis of a dry cask is essential to assess the integrity of the system itself, and of the SNF stored within it. In this regard, both heat production and heat removal mechanisms have to be taken into account. On this basis, the present paper addresses the variation in the system heat removal capacity when considering different backfill gas pressures. In particular, the analysis, carried out with the MELCOR code, investigates the thermal response of the ventilated, concrete-based HI-STORM 100S cask, currently employed for spent fuel elements of Light Water Reactors (LWRs), when imposing different initial pressures for the helium backfill gas. Results are reported primarily in terms of maximum temperature of the fuel cladding, which is the variable under regulatory surveillance. In addition, the adherence to the maximum design pressure for the canister is verified by evaluating the helium pressure as the steady state is reached. The analysis seems to suggest that the safe operation of the HI-STORM 100S cask is guaranteed only for a limited range of the initial helium pressure. Full article

This study investigated how welding affects the thermal deformation of square cells produced for casks, which are dry storage containers for spent nuclear fuel. We aimed to minimize structural deformation by utilizing STS316L as the material for the square cells. We explored a method of subdividing the square cells and joining them through butt welding. Keeping the upper plate thickness constant, GTA butt welding was conducted while varying the column’s wall thickness, followed by measurement with a laser vision sensor. The heat conduction and thermal strain were then calculated using a finite element analysis (FEM). Both experimental and analytical results confirmed that there was significant thermal deformation in the cases of thick-walled columns due to variations in heat conduction distribution, with the resulting deformation patterns depending on thickness. Full article

This limited-scope study demonstrates the application of probabilistic risk assessment (PRA) methodologies to a spent fuel storage system for spent pebble-filled dry cask with a focus only on the necessary PRA technical elements sufficient to risk-inform the spent fuel storage system design. A dropping canister scenario in a silo of the spent fuel storage system is analyzed through an initiating event (IE) identification from the Master Logic Diagram (MLD); event sequence analysis (ES) by establishing the event tree; data analysis (DA) for event sequence quantification (ESQ) with uncertainty quantification; mechanistic source term (MST) analysis by using ORIGEN; radiological consequence analysis (RC) by deploying MicroShield, and risk integration (RI) by showing the Frequency-Consequence (F-C) target curve in the emergency area boundary (EAB). Additionally, a sensitivity study is conducted using the ordinary least square (OLS) regression method to assess the impact of variables such as failed pebble numbers, their location in the canister, and building wall thickness. Furthermore, the release categories grouped from the end states in the event tree are verified as safety cases through the F-C curve. This study highlights the implementation of PRA elements in a logical and structured manner, using appropriate methodologies and computational tools, thereby showing how to risk-inform the design of a dry cask system for storing spent pebble-filled fuel. Full article

Detecting helium leakage is important in many applications, such as in dry cask nuclear waste storage systems. This work develops a helium detection system based on the relative permittivity (dielectric constant) difference between air and helium. This difference changes the status of an electrostatic microelectromechanical system (MEMS) switch. The switch is a capacitive-based device and requires a very negligible amount of power. Exciting the switch’s electrical resonance enhances the MEMS switch sensitivity to detect low helium concentration. This work simulates two different MEMS switch configurations: a cantilever-based MEMS modeled as a single-degree-freedom model and a clamped-clamped beam MEMS molded using the COMSOL Multiphysics finite-element software. While both configurations demonstrate the switch’s simple operation concept, the clamped-clamped beam was selected for detailed parametric characterization due to its comprehensive modeling approach. The beam detects at least 5% helium concentration levels when excited at 3.8 MHz, near electrical resonance. The switch performance decreases at lower excitation frequencies or increases the circuit resistance. The MEMS sensor detection level was relatively immune to beam thickness and parasitic capacitance changes. However, higher parasitic capacitance increases the switch’s susceptibility to errors, fluctuations, and uncertainties. Full article

The operating license of Kori units 3 and 4 are to be expired in 2024 and 2025. If the plants are decided to be decommissioned, the spent nuclear fuels in the spent fuel storage pool (SFP) have to be removed from the site. However, no proper storage facility is currently available in Korea. To overcome the difficulty, this study proposes application of an independent temporary SFP cooling system. It is expected to safely isolate the SFP from the plant. The case study conducted showed this concept is achievable and technically mature. Moreover, the installation cost per unit is USD 9.51 million obtained from an analogous estimating. This is much lower cost than the dry cask options. Then, system requirements and applicable design are developed. Maximum heat generation of unit 4 is estimated as 1.67 MW, through fuel characteristics categorizations and ORIGEN calculations. To remove the decay heat, installation of an indirect air-cooled method is selected by evaluating three cooling methods. The modification includes new heat exchanger, secondary cooling pumps, and chillers. Furthermore, two single failures are considered; cooling pump and normal power. This can be covered by two redundant trains and a back-up diesel generator provided. Full article

Nuclear safety relies to a good extent on thoroughly validated codes. However, code predictions are affected by uncertainties that need to be quantified for a more accurate evaluation of safety margins. In this regard, the present paper proposes a preliminary uncertainty and sensitivity analysis of the thermal behavior of a concrete-based dry cask for spent nuclear fuel storage, employing the MELCOR code and a series of MATLAB scripts. As thermal behavior is of utmost importance for the fulfillment of United States Nuclear Regulatory Commission (USNRC) safety requirements, the Peak Cladding Temperature (PCT) has been addressed as the key Figure of Merit (FOM). Variables related to the main heat transfer mechanisms have been selected as input parameters for the uncertainty quantification, whereas heat source and heat sink, namely decay power and external air temperature, have been dealt with in a separate sensitivity analysis. The results show that the selected parameters have a weak influence on the PCT, whereas it is strongly related to the decay power and external air temperature values. In any case, PCT stays below the regulatory threshold even under the considered off-normal conditions. Full article

A key issue contributing to the success of NPP technology is the safe handling of radioactive waste, particularly spent nuclear fuel. According to the IAEA safety standard, the spent fuel must be stored in interim wet storage for several years so the radiation and the decay heat of the spent fuel will decrease to the safe limit values, after which the spent fuel can be moved to dry storage. In this study, we performed a theoretical analysis of heat removal by natural convection airflow in spent nuclear fuel dry storage. The temperature difference between the air inside and outside dry storage produces an air density difference. The air density difference causes a pressure difference, which then generates natural airflow. The result of the theoretical analysis was validated with simulation software and experimental investigation using a reduced-scale dry storage prototype. The dry storage prototype consisted of a dry cask body and two canisters stacked to store materials testing reactor (MTR) spent fuel, which generates decay heat. The cask body had four air inlet vents on the bottom and four air outlet vents at the top. To simulate the decay heat from the spent fuel in the two canisters, the canisters were wrapped with an electric wire heater that was connected to a voltage regulator to adjust the heat power. The theoretical analysis results of this study are relatively consistent with the experimental results, with the mean relative deviation (MRD) values for the prediction of air velocity, the heat rate using natural airflow, and the heat rate using the thermal resistance network equation are +0.76, −23.69, and −29.54%, respectively. Full article

The paper assumes that, at the end of the operational period of a Spanish nuclear power plant, an Independent Spent Fuel Storage Installation will be used for long-term storage. Spent fuel assemblies are selected and transferred to casks for dry storage, with a series of imposed restrictions (e.g., limiting the thermal load). In this context, we present a variant of the problem of spent nuclear fuel cask loading in one stage (i.e., the fuel is completely transferred from the spent fuel pool to the casks at once), offering a multi-start metaheuristic of three phases. (1) A mixed integer linear programming (MILP-1) model is used to minimize the cost of the casks required. (2) A deterministic algorithm (A1) assigns the spent fuel assemblies to a specific region of a specific cask based on an MILP-1 solution. (3) Starting from the A1 solutions, a local search algorithm (A2) minimizes the standard deviation of the thermal load among casks. Instances with 1200 fuel assemblies (and six intervals for the decay heat) are optimally solved by MILP-1 plus A1 in less than one second. Additionally, A2 gets a Pearson’s coefficient of variation lower than 0.75% in less than 260s CPU (1000 iterations). Full article

In the dry cask storage of spent nuclear fuels, a stainless-steel canister acts as an important barrier for encapsulating spent fuels. As a result, local corrosion behavior of 304L stainless steel has become an issue of concern in the wet coastal region and salt spray environment. The test was conducted after deposition of simulated sea salt particles on the 304L stainless-steel specimen. It was first covered with a crevice former, and then kept at 45 °C with a relative humidity of 45%, 55%, and 70%, respectively. The surface morphologies and electron back scatter diffraction (EBSD) analysis of the corroded region for the 304L stainless-steel specimen are presented in this paper. The goal of this work was to investigate the crevice corrosion behavior of 304L stainless steel under different chloride concentrations and relative humidity conditions. From the experimental results, a threshold relative humidity for stress corrosion cracking (SCC) initiation of 304L stainless steel was proposed. Full article

Elevated temperature, gamma radiation, and geometric constraints inside dry storage casks for spent nuclear fuel represent a harsh environment for nondestructive inspection of the cask and require that the inspection be conducted with a robotic system. Electromagnetic acoustic transducers (EMATs) using non-contact ultrasonic transduction based on the Lorentz force to excite/receive ultrasonic waves are suited for use in the robotic inspection. Periodic permanent magnet EMATs that actuate/receive shear horizontal guided waves are developed for application to robotic nondestructive inspection of stress corrosion cracks in the heat affected zone of welds in stainless steel dry storage canisters. The EMAT’s components are carefully selected in consideration of the inspection environment, and tested under elevated temperature and gamma radiation doses up to 177 °C and 5920 krad, respectively, to evaluate the performance of the EMATs under realistic environmental conditions. The effect of gamma radiation is minimal, but the EMAT’s performance is affected by temperatures above 121 °C due to the low Curie temperature of the magnets. Different magnets are needed to operate at 177 °C. The EMAT’s capability to detect notches is also evaluated from B-scan measurements on 304 stainless steel welded plate containing surface-breaking notches. Full article

In this study, the heat flow characteristics and thermal performance of a dry storage cask were investigated via thermal flow experiments and a computational fluid dynamics (CFD) simulation. The results indicate that there are many inner circulations in the flow channel of the cask (the channel width is 10 cm). These circulations affect the channel airflow efficiency, which in turn affects the heat dissipation of the dry storage cask. The daily operating temperatures at the top concrete lid and the upper locations of the concrete cask are higher than those permitted by the design specification. The installation of the salt particle collection device has a limited negative effect on the thermal dissipation performance of the dry storage cask. Full article

The microstructure of Al/B₄C metal matrix composites (MMCs) used as neutron absorbers in both dry storage casks and wet storage pools of spent nuclear fuel was analyzed by SEM and TEM. A polishing method of a focused Ga⁺ ion beam was used to obtain an ideal sample surface with very low roughness, which was used to statistically analyze the distribution characteristics and size factor of B₄C particles in the aluminum matrix. The area of B₄C particles mainly ranged from 0 to 0.5 μm², which was the proportion of 64.29%, 86.99% and 76.86% of total statistical results for the Al-15%B₄C, Al-25%B₄C and Al-30%B₄C MMCs, respectively. The average area of B₄C particles in the Al-15%B₄C, Al-25%B₄C and Al-30%B₄C MMCs were about 1.396, 0.528 and 1.183 μm², respectively. The nanoscale precipitates were analyzed by the element mappings in scanning transmission electron microscopy (STEM) mode and electron energy loss spectroscopy (EELS) mode, which included elliptic alloy precipitates with elemental Cu, Cr, Fe and Si, except for Al, and B₄C nanoparticles with polygonal shape. The interface characteristics showed that the (021) crystal plane of B₄C particle and (111) crystal plane of aluminum matrix grew together. The lattice misfit was about 1.68% for (111)_Al//(021)_B4C. The corrosion properties and corresponding mechanism of Al/B₄C MMCs were investigated in an aqueous solution with 5000 ppm boric acid at 100 °C and atmospheric pressure, which showed that the mass increment rate was first decreased with increasing corrosion time and then increased. Full article