A high-burnup fuel management strategy to extend the fuel lifetime of spent high-temperature gas-cooled reactor (HTGR) fuel for a once-through fuel cycle had been proposed in previous research. The strategy is able to increase fuel utilization efficiency without compromising the safety and proliferation resistance features of the tristructural isotropic particle fuel. This study aims to provide quantitative insights into the potential effects of core design changes aimed to implement the strategy to directly reuse spent fuel, focusing on thermal-hydraulic performance under both steady-state operation and accident scenarios. Neutronics and thermal-hydraulic analysis have been performed using a sample core layout for the direct reuse of spent fuel. This study puts a heavier focus on the thermal-hydraulic analysis, with the neutronics analysis as a supplement to highlight the trade-offs between enhanced fuel utilization and thermal safety performance associated with the direct reuse of spent fuel in a two-region HTGR core. Results showed that simple modifications could be done to a reference HTGR core design to form a two-region core to directly reuse spent fuel assemblies without significant degradation in core performance. The fuel burnup could be increased by 10% with this design without deterioration in core safety parameters in terms of reactor kinetics. The steady-state thermal-hydraulic analysis with a simplified RELAP5-3D model showed that a two-region core could also operate at fuel temperatures similar to its reference core, which could be achieved by having a coolant flow distribution based on the power ratio between the two fuel regions. The core modifications only caused the peak fuel temperature during a depressurized loss-of–forced cooling scenario to increase by approximately 60°C higher compared to its reference core.

To improve the safety of road transportation of Spent Nuclear Fuel (SNF), this paper proposes a novel approach for risk identification and chaotic synchronous control in SNF road transportation systems. Firstly, a dynamic risk evolution model for the road transportation of SNF is developed by analyzing the nonlinear interactions among vehicles, environmental conditions, and human factors using complex network analysis and nonlinear dynamics. Secondly, an enhanced K-shell decomposition method is applied to identify key risk nodes and assess the relative importance of different risk factors, providing a basis for targeted risk control. Finally, a chaotic synchronization control strategy based on Lyapunov stability is proposed to suppress risk divergence and restore system stability. Three targeted control schemes are evaluated by varying the control gain coefficients across the ‘Vehicle–Environment–Human’ dimensions. Simulation results indicate that the strategy prioritizing environmental and human risk control yields the fastest convergence, significantly outperforming vehicle-centric approaches. The results show that prioritizing both environmental and human-factor control is most effective for suppressing chaotic divergence. This provides a solid quantitative basis for the strategic shift from passive defense to active environmental warning, thereby significantly optimizing the dynamic risk management of the SNF transportation system.

In June of 2023, the U.S. Department of Energy (DOE), Office of Nuclear Energy (NE) announced the selection of its Collaboration-Based Siting (CBS) Consortia, a group of 12 awardees, including the Consortium for Risk Evaluation with Stakeholder Participation (CRESP) (led by Vanderbilt University) to assist DOE-NE with the development of its process for siting a federal consolidated interim storage facility (FCISF) for spent nuclear fuel (SNF) storage. At this time, DOE NE is not soliciting interested host communities, rather the CBS Consortia are tasked with in-depth engagement, mutual learning, and capacity building to provide DOE NE with feedback on the CBS process. CRESP’s objective is to engage communities in two regions (the Pacific Northwest and the Southeast) with sites currently storing defense- and research-related SNF to foster learning concerning the best and worst practices in community participation in risk-informed decision making. This includes learning from existing structures for public input in radioactive waste management decision making [e.g. citizen advisory boards (CABs)] and other local parties about how to build and sustain trust among the parties. CRESP has been working to engage stakeholders and Tribes surrounding two DOE sites historically differing in receptiveness to engaging with DOE and trusting in DOE to accomplish its missions. CRESP’s approach to date has consisted of (1) engaging voluntarily members and former members of the DOE Office of Environmental Management’s CABs to develop a mutual understanding of their perspectives, values, and experiences related to risk and participatory decision making; (2) engaging members of those communities that would likely be part of any radioactive waste management discussions and who may provide valuable feedback for the development of the CBS process; and in the longer term, (3) engaging communities to foster knowledge sharing on understanding and definitions of risk and how risk is factored into community decision making. Our emphasis is to identify opportunities for improving risk communication frameworks, strategies, and decision making. The selection of a future FCISF site is likely to result in the creation of a structure similar to a CAB composed of members of the local community. We anticipate that these insights can help DOE NE learn from existing participatory risk communication structures in place at sites storing defense- or research-related SNF to understand best practices for fostering enduring and participatory relationships with future CABs. Within this paper, we (1) describe CRESP’s overall approach to assisting DOE NE with maturing the CBS process; (2) present a summary of preliminary phase 1 project results, including the development of a body of knowledge (describing available resources to support community engagement, risk communication, and participatory decision making) and community ecosystem information (leveraging demographic, social, economic, and environmental attribute mapping and advanced sentiment analysis of social media data); and (3) based on these results, provide observations for future research opportunities to support the development of CBS processes for radioactive waste management facilities, generally.

Long-Term Corrosion Behavior of a Cold-Spray Copper-SiC Composite Coating for Spent Nuclear Fuel Disposal Canisters

Floating Nuclear Power Plants (FNPPs) present a novel approach to energy production, offering flexible and resilient power solutions for remote and coastal regions. However, their deployment raises significant legal, environmental, and security concerns under international law. This article critically examines whether existing legal frameworks adequately regulate FNPPs or if international law reform is necessary to address emerging risks. Key areas of analysis include nuclear safety, liability for accidents, environmental protection under the United Nations Convention on the Law of the Sea (UNCLOS), and potential security threats such as nuclear proliferation and maritime terrorism. The study explores gaps in current treaties, including the Convention on Nuclear Safety and the Joint Convention on the Safety of Spent Fuel Management, highlighting challenges posed by FNPP mobility and jurisdictional ambiguities. By assessing state responsibilities and international obligations, the article argues for targeted legal reforms to enhance regulatory clarity, safety, and accountability in the governance of FNPPs.

The study addresses the assessment and management of environmental safety at spent nuclear fuel (SNF) storage facilities under emergency conditions. The research aims to improve environmental safety management by integrating modern risk assessment approaches, external threat analysis, and advanced modeling tools. The focus is on understanding the processes of radiation formation, propagation, and deposition during accidents at SNF storage facilities, as well as on developing methods for evaluating contamination levels and the dispersion of radionuclides in the surrounding environment. A detailed sequence of procedures for radiation impact assessment is proposed, beginning with the collection of comprehensive data on the emission source, including radionuclide composition, activity levels, release height,speed, and direction. Meteorological data such as wind speed and direction, temperature, humidity, and atmospheric stability are incorporated, alongside topographic characteristics, which together determine the spatial and temporal spread of radioactive contamination. Diffusion models, particularly the Gaussian model, are employed to predict concentrations of radioactive admixtures, evaluate horizontal and vertical distributions in the atmosphere, describe dry deposition of particles, and estimate soil contamination. Mathematical relationships are presented for calculating radionuclide concentrations, dispersion coefficients, surface contamination density, and radiation doses to both personnel and the general population. The study also critically analyzes the limitations of the Gaussian model, including its assumptions of atmospheric homogeneity, simplified treatment of deposition processes, and lack of corrections for complex terrain, urban structures, convective flows, and variable turbulence. To address these limitations, complementary approaches, such as Lagrangian particle models and computational fluid dynamics (CFD) simulations, are discussed to provide more accurate predictions under heterogeneous and dynamic conditions. Emphasis is placed on integrating these methods with statistical analysis of particle size distributions, deposition rates, and emission scenarios to enhance reliability and applicability in real-world conditions. Based on the modeling results, a comprehensive framework for radiation risk assessment is developed, enabling the calculation of external and internal exposure through integral formulas and the estimation of the probability of adverse health effects. The proposed methodology allows for quantitative evaluation of risks that exceed permissible exposure levels and supports timely decision-making regarding protective measures. The findings can be applied to enhance environmental monitoring systems, improve the accuracy of predicting radioactive impact zones, and optimize the management of SNF storage facilities during emergency scenarios, ultimately contributing to increased environmental and public safety, as well as the development of guidelines for emergency preparedness and response.

Adsorptive separation of xenon (Xe) and krypton (Kr) is a promising yet challenging technique due to their similar physical properties. In this work, we present a microporous metal-organic framework (MOF) with open metal sites exposed on the small caged pores for efficient Xe/Kr separation. The cage-based MOF, Cu-BTA, features well-defined permanent pores and exhibits record-high Xe gravimetric and exceptionally high volumetric uptake capacities of 7.61 mmol g-1 and 7.54 mmol cm-3 at 298 K and 100 kPa. Breakthrough experiments confirm its excellent separation performance for 20:80 Xe/Kr mixtures under both dry and highly humid conditions with the highest dynamic Xe uptake of 4.92 mmol g-1. Moreover, the material demonstrates efficient Xe capture even at an ultralow concentration (400 ppm), suggesting its great potential for removing radioactive Xe from used nuclear fuel (UNF) reprocessing off-gas.

The effective neutron multiplication factor (<em>k<sub>eff</sub></em>) of a fissile system is a function of space and time f(x, y, z, t), where (x, y, z) are the coordinates of the reference point in space, and <em>t</em> is time. The evidence of the time-dependence of the neutron multiplication factor of a fissile system was published in 2017. The spatial dependence is as important as the time dependence, as interacting arrays of fissile material during storage can lead to an increase in the system (<em>k<sub>eff</sub></em>). Therefore, evaluation of how the storage matrix influences the <em>k<sub>eff</sub></em> should be an integral part of the safety analysis in the design of the spent fuel storage facility. The selection of an appropriate storage matrix must be based on and supported by safety analysis and should precede any storage of fissile material—especially fresh fuel and spent fuel in which no credit for burnup is taken into account. In addition, the selected storage configuration must be approved by the nuclear regulatory authorities, as an improper arrangement may increase the <em>k<sub>eff</sub></em> beyond the regulatory limit and result in criticality accidents. Since the time dependence of the neutron multiplication factor has already been performed, this research will focus on investigating how the spatial arrangement of spent fuel in spent fuel storage can influence the <em>k<sub>eff</sub></em>.

An essential early phase of any new nuclear power plant (NPP) project is drafting and preparation of a radioactive waste management strategy. The paper outlines major steps in such strategy presented in a case study approach for new NPP project in Krško (the JEK2 project). The paper includes analyses of options and strategies for radioactive waste (RAW) and spent nuclear fuel (SNF) management. This article presents a proposed management plan for RAW and SNF expected to be produced during operation and decommissioning of JEK2, together with an assessment of anticipated decommissioning and waste management costs. The document further examines the existing financing framework for radioactive waste management in Slovenia, which is legally regulated for current nuclear facilities. A similar approach is planned for JEK2, including the expansion of existing and planned facilities for waste handling and storage. Accordingly, additional financial resources will be required through contributions to the national decommissioning and disposal fund. An analysis was therefore conducted to estimate the scale of additional contributions needed to ensure adequate funding for new NPP project in Krško.

The paper provides a generalized overview of the components of the nuclear fuel cycle (NFC) in Ukraine and worldwide, as well as specific features of open and closed fuel cycle configurations. Using the Nuclear Fuel Cycle Simulation System (NFCSS) developed by the International Atomic Energy Agency (IAEA), an assessment was performed to model nuclear material flows during VVER-1000 operation under two scenarios: use of uranium oxide fuel (open fuel cycle) and partial core loading with mixed uranium-plutonium oxide (MOX) fuel derived from spent nuclear fuel (SNF) reprocessing (closed fuel cycle). The impact of SNF reprocessing for natural uranium, enrichment requirements and waste generation was evaluated, as well as changes in isotopic composition, radiotoxicity and residual heat release. It was demonstrated that the use of MOX fuel can reduce natural uranium consumption by approximately 20–25%, decrease the scope of enrichment activities and enable more efficient utilization of available nuclear materials. At the same time, the total mass of SNF remains nearly unchanged in both scenarios, although its isotopic composition undergoes significant transformation. The calculations revealed an increase in residual heat generation and overall radiotoxicity of spent nuclear fuel in the scenario involving MOX fuel, which introduces additional engineering challenges for the design of storage and geological disposal systems for high-level radioactive waste, particularly regarding thermal loads and long-term radiological impact. The study highlights NFCSS feasibility as an effective tool for strategic analysis of nuclear fuel cycle development scenarios, nuclear fuel utilization planning and evaluation of SNF reprocessing efficiency to support the improvement of nuclear industry in Ukraine.

Recently, the search for accident-tolerant nuclear fuel has become vital in light of the increasing global demand for energy and the trend of many countries to rely on nuclear energy as an environmentally friendly energy source. Therefore, this study investigates the feasibility and performance of various advanced fuel types, including uranium carbide, uranium nitride, and thorium-uranium–based carbides and nitrides [(Th, 233U)C, (Th, 233U)Nnat, and (Th, 233U)15], as alternatives to conventional uranium carbide oxide fuel for small modular advanced high-temperature reactors (SmAHTR). A three-dimensional full-core model of a SmAHTR was developed using the MCNPX 2.7 code to conduct a comprehensive neutronic analysis over an irradiation period of 1450 effective full-power days. The key parameters evaluated included the effective multiplication factor, fissile inventory ratio, production of plutonium and minor actinides, and the evolution of radioactivity in actinide and nonactinide isotopes. The results showed that thorium-uranium–based carbides and nitrides offer superior reactivity, longer fuel cycles (up to a 45% increase), and a complete elimination of reactor-grade plutonium and minor actinide formation. The suggested fuels exhibited favorable safety characteristics. Additionally, the study provided detailed insights into radial power and neutron flux distribution, confirming the absence of hot spots. Among the examined options, (Th, 233U)15N emerged as the most promising accident-tolerant fuel for SmAHTRs in terms of both performance and safety. These findings support the development of thorium-based nitride fuels as viable candidates for next-generation small modular reactors.

This study presents neutron-physical modeling of radionuclide accumulation in a VVER-1000 reactor after the first fuel cycle using the MCNP6 code and the ENDF/B-VII nuclear data libraries. Calculations were performed to determine the generation of major long-lived fission products (⁹⁰Sr, ⁹⁹Tc, ¹³⁷Cs, ¹²⁹I, etc.), actinides (Np, Pu, Am), noble gases (Kr, Xe), as well as aerosol-forming and iodine-containing radionuclides. Special attention is given to the evaluation of residual activity, and the impact of accumulated isotopes on radiation safety. The findings are of practical importance for the development of spent nuclear fuel (SNF) management strategies, storage design, environmental risk assessment, and preparation for nuclear power plant construction in Kazakhstan.

Abstract It is well known that significant amounts of radioxenon radionuclides are released from Medical Isotope Production Facilities and to a lesser extent from Nuclear Power Plants (NPP). These emissions cause a background in the atmosphere that is often detected by noble gas systems of the International Monitoring System (IMS) operated by the Comprehensive Nuclear-Test-Ban Treaty Organization Preparatory Commission for nuclear explosion monitoring. In addition to those facilities, the operation of a Spent Nuclear Fuel (SNF) reprocessing plant may possibly also contribute to the IMS observations, but this has not yet been investigated. Even after long cooling time, the short-lived radioxenon isotopes are present in spent fuel due to spontaneous fission with the isotopes 244 Cm and 240 Pu being the main contributors. The SNF reprocessing process can promptly release the whole radioxenon inventory if there is no retention system. The aim of this work is to investigate the possible radioxenon emission during SNF reprocessing caused by spontaneous fission of heavy elements. Two independent methods are applied to determine the radioxenon releases. One approach is to use the published release of 131 I as a proxy. The other is to analyse the parameters of reprocessed spent fuel to determine the content of 244 Cm and 240 Pu and with this information to estimate the radioxenon inventory. It turns out that the estimated maximum release of 133 Xe is of the order of GBq/day which is almost as high as the average discharge on an NPP site. Assuming the absence of an effective retention system that prevents the release of radioxenon into the environment, the results of the calculations show that industrial scale reprocessing plants should be considered as a weak but not negligible source of radioxenon.

Preliminary Test and Analysis of TWIN-Tandem Wire Arc Additive Manufacturing for Producing High-Purity Copper Overpacks in Spent Nuclear Fuel Disposal Canisters: Focus on Microstructure and Dilution Behavior

Potential Applications of Artificial Intelligence in Spent Nuclear Fuel Reprocessing Research: A Review

This study explores the possibility of the direct reuse of spent fuel from the NuScale small modular reactor (SMR) 160-MW(thermal) in CANDU6 reactors via the DUPIC (Direct Use of spent PWR fuel in CANDU reactors) cycle, employing the Oxidation and Reduction of Oxide (OREOX) process to mitigate nuclear proliferation risks and improve spent fuel management. The analysis begins with the isotopic inventory of SMR spent fuel after a fuel burnup of 60 GWd/tonne heavy metals (HM) followed by 5 years of cooling. Neutronics and thermal-hydraulic simulations were performed on two OREOX fuel configurations, both utilizing D2O as the moderator, with the difference in the coolant materials being D2O in the first configuration and H2O in the second. The neutronics analysis was carried out using a full-core model. Based on the resulting power and flux profiles, the thermal-hydraulic analysis was then focused on the hottest pressure tube. The neutronics results showed that OREOX fuel increases the inventory of 238Pu, hindering the reprocessing of spent fuel for weapons-grade plutonium. Additionally, the OREOX process significantly depletes transuranic elements such as 239Pu, 241Pu, 241Am, and 243Am, further reducing proliferation risks. The configuration with D2O as both coolant and moderator demonstrated the longest fuel cycle (300 effective full-power days (EFPDs)) with a burnup of 7.26 GWd/tonne HM, along with an improved power distribution in comparison to both standard UO2 fuel and the second OREOX configuration. The thermal-hydraulic results further confirmed the safe operation of the hottest pressure tube in this configuration, showing higher safety margins than the alternative, while considering the substantial decrease in the thermophysical properties of the OREOX fuel resulting from burnup in SMRs.

Recycling used nuclear fuel (UNF) can be accomplished through the high temperature, electrochemical molten salt electrolyte process known as pyroprocessing. In the first electrochemical step of pyroprocessing, ceramic UNF is reduced to its metallic state in a molten LiCl-Li2O salt. To counterbalance the oxide reduction, oxygen is evolved at an anode. Ruthenium was recently discovering as an effective anode in this process with it experiencing minimal corrosion during operation. However, as a precious metal, the acquisition of ruthenium represents a large cost for pyroprocessing. One method to reduce the cost of the anode is by producing a ruthenium alloy that includes a cost-effective alloying element. In this study, ruthenium alloyed with Fe, Mo, and Ni were investigated as potential alternatives to pure ruthenium as an anode. Electrochemical and surface characterization results will be discussed. Additionally, a pretreatment method that can improve the performance of anodes will be presented.This work was performed under ARPA-e grant DE-AR0001697. XPS was purchased under NSF MRI award 2117820. C.M. is supported by DOE Fellowship under award DE-NE0009105.

Extensive research on the electrochemistry of actinides in molten chloride and fluoride salts has been conducted at the Joint Research Center Karlsruhe (JRC) since the early 2000s, with the aim of supporting the development of electrochemical processes for the recovery of actinides from spent nuclear fuel and for studying the chemistry and fuel cycle of molten salt reactors. The investigations rely on standard transient electrochemical techniques, such as cyclic voltammetry and chronopotentiometry. However, the use of molten salt media at high temperatures requires specialized experimental equipment, particularly a pure inert atmosphere to protect highly hygroscopic materials from unwanted reactions with moisture and oxygen. On the other hand, molten salts are highly efficient electrolytes due to their high conductivity and wide electrochemical window. Additionally, elevated working temperatures enable fast electrode diffusion and reaction kinetics.The experimental setup for electrochemical measurements in molten salt media, developed and installed at JRC Karlsruhe, consists of a high-temperature Inconel electrolyzer housed in a glove box maintained under purified argon to keep oxygen and moisture levels below 2 ppm. A hydrogen fluoride gas line is connected to the glove box, allowing for purification of fluoride melts by bubbling pure HF gas. This, combined with the ability to work with gram-scale quantities of actinides, makes the setup unique. The setup is described in detail in [1]. Chloride melts typically do not require purification, as they are commercially available in sufficient quality, dehydrated, and packed under pure argon.The present work describes this experimental setup and summarizes key results from basic electrochemical studies of actinides, as well as the development of an electrorefining process for the recovery of actinides from metallic spent nuclear fuel. In the basic studies, information was obtained on the electrochemical behavior of actinides (redox and deposition potentials, reaction mechanisms), thermochemical data (activity coefficients, enthalpies, and entropies of formation), and salt transport parameters (diffusion coefficients). In LiCl-KCl eutectic melt, the electrochemical properties of Th, U, Np, Pu, and Am were measured using both inert (W, Mo) and reactive (Bi, Cd, Ni, Al) working electrodes. References to these measurements can be found in [2]. Electrochemical studies in fluoride melts started at a later stage and were focused on developing an efficient purification method, as well as verifying the purity and characterizing the behavior of Th in LiF-CaF2 eutectic melt [1], along with unpublished data on U in the same system.The use of reactive electrodes proved particularly advantageous for actinide recovery, as the metals reduced on such cathodes form stable alloys with the electrode material, thereby preventing undesirable side reactions of the deposited species. However, the selective recovery of individual actinides is limited when using reactive cathodes, due to smaller differences in their reduction potentials compared to inert electrodes. A comparison of the apparent standard potentials of selected actinides on various electrode materials is shown in Fig. 1. Nonetheless, group-selective separation of actinides from lanthanides is possible, and this has been effectively applied in the development of an electrorefining process for homogeneous recovery of all actinides from metallic spent nuclear fuel using solid reactive aluminum cathodes [3]. Fig. 1 Apparent electrochemical potentials measured for various Ln and An on solid W, Al and liquid Cd working electrodes in a LiCl-KCl eutectic melt at 450°C. In this process, actinides are electro-separated from fission products in a LiCl-KCl eutectic molten salt at 450 °C by applying a constant current between the metallic fuel, placed in a tantalum basket, and an aluminum cathode. During electrolysis, actinide cations, generated via anodic oxidation of the fuel, are transported through the electrolyte and deposited onto the aluminum cathode. Meanwhile, alkali metals, alkaline earth metals, and rare earth fission products are also dissolved into the melt, but are not reduced due to the controlled deposition potential. The selectivity and efficiency of this process have been demonstrated by electrorefining both non-irradiated and irradiated An-Ln-Zr fuel produced at JRC Karlsruhe (METAPHIX-1, composition U61-Pu22-Am2-Nd3.5-Gd0.5-Y0.5-Ce0.5-Zr10) [4]. Souček, P., et al., Electrochimica Acta, 2021. 380: p. 138198.Souček, P. and R. Malmbeck, 17 - Pyrochemical processes for recovery of actinides from spent nuclear fuels, in Reprocessing and Recycling of Spent Nuclear Fuel, R. Taylor, Editor. 2015, Woodhead Publishing: Oxford. p. 437-456.Souček, P., et al., Journal of Nuclear Materials, 2015. 459(0): p. 114-121.Ohta, H., et al., Journal of Nuclear Science and Technology, 2011. 48(4): p. 654-661. Figure 1

Accurate radiation safety evaluation is essential for the safe transportation and storage of spent nuclear fuel. Conventional approaches, which rely primarily on surface dose measurements or simplified models, often fail to account for the complex particle composition, energy spectra, and spatial flux distribution of highly radioactive spent fuel modules. This study employs the Geant4 Monte Carlo simulation toolkit to analyse the shielding performance of composite materials designed for spent nuclear fuel casks. We validate Geant4’s gamma attenuation predictions against EpiXS data in the energy range 0.1-20 MeV, comparing linear attenuation coefficients and half-value layer (HVL) values for candidate materials. Simulations are performed under environmental conditions representative of Rooppur Nuclear Power Plant in Bangladesh. The results show that the composite designs achieve shielding performance comparable to or superior to conventional materials, with improved reliability and optimized weight. Geant4 predictions closely match EpiXS data across tested energies, demonstrating Geant4’s suitability for detailed radiation safety evaluations in spent nuclear fuel cask design. These findings guide advanced, weight-efficient shielding strategies to enhance safety and cost-effectiveness in nuclear fuel storage and transportation. Deviations between Geant4 and EpiXS results remained within ±2%, confirming strong computational accuracy and supporting the design’s potential for enhanced shielding performance and mass optimization.

Assessing spent nuclear fuel (SNF) integrity is essential for ensuring radiation safety throughout its storage duration. One of the widely used methods is the sipping test, a nondestructive technique for detecting leaks in SNF by analyzing the release of radioactive material. While sipping tests have been applied to various reactor types, detailed designs and procedures for SNF from the Materials Test Reactor (MTR) remain limited. This study presents the design and operational procedure of a sipping test for MTR-type SNF from the Gerrit Augustinus Siwabessy research reactor in a wet storage facility, covering key components and procedures for the test. The test system is comprised of a custom-designed sipping tube, SNF basket, crane, and supply pipes for demineralized water and compressed air. The results demonstrate efficient system functionality, ensuring safe and smooth SNF handling, effective pool water replacement by demineralized water injection, thorough homogenization before sampling, and reliable test water sampling over multiple test intervals. The findings provide insights into the effectiveness of the sipping test for SNF integrity assessment and contribute to best practices for the management of a wet storage facility.