Reflooding is a key process in the cooling of debris beds. To better investigate the behavioral characteristics during debris bed reflooding, this study establishes a porous media non-thermal equilibrium model suitable for homogeneous debris beds based on the CFD method. The heat transfer correlations are defined for single-phase vapor region, film boiling region, nucleate boiling region, and single-phase liquid region. By coupling with the multiphase flow model, accurate simulations are achieved for bottom and top injection reflooding processes in terms of reflooding time, quench front propagation, and pressure. Furthermore, the reflooding progression and flow field of the debris bed are analyzed. Bottom injection reflooding exhibits strong one-dimensional characteristics, but the quench front shows significant disturbances and an uneven flow field distribution. In top injection reflooding, the quench front center features a downward flow velocity, while near-wall regions display upward flow velocity, with other regions showing flow directed radially toward the wall. This study provides the theoretical foundation and technical support for understanding debris bed reflooding behavior and optimizing debris bed cooling strategies.
This work presents a design concept for heat pipe cooled microreactors by proposing a liquid lead bond as an alternative to the conventional stainless steel monolithic core. A transient heat pipe failure analysis is conducted to assess the passive thermal resilience and safety characteristics of this novel design. A three-dimensional finite element model was developed in ANSYS to perform transient simulations under postulated heat pipe failure conditions, including single, peripheral, and cascade scenarios. The transient results show that the reactor exhibits rapid but self-limiting temperature increases immediately after failure, reaching a new quasi-steady state within 400-500 s. The cascade failure produced the most severe condition, with the maximum fuel temperature rising to approximately 1260 K (about 220 K above steady-state), which remains well below the uranium nitride fuel melting point (3120 K). Under the modeled conditions, the lead-bonded design shows a 35-40% reduction in peak temperature overshoot, with the normalized temperature ratio decreasing from 1.38 to 1.21. The fuel safety margin increases from 46% to 55%. Also, the liquid lead bond shortens the stabilization time and smooths radial temperature gradients, thereby mitigating thermal stresses during transients. These results suggest that the liquid lead-bonded configuration may offer enhanced thermal margins and transient tolerance relative to conventional stainless steel cores. Future work will extend this framework to coupled neutronic-thermal analysis to evaluate reactivity feedback and core dynamic stability.
A set of modeling improvements is developed and tested in the three-dimensional module of the system code CATHARE3 to enhance the prediction of single- and two-phase turbulent mixing in rod bundle sub-channel configurations. The new formulation includes a porosity-consistent Reynolds stress tensor, a data-driven correlation for the turbulent and dispersive viscosity, a revised thermal diffusivity based on a double-averaging framework, and a regime-dependent void-dispersion model. These developments are assessed against the Rowe and Angle 1967 experimental database, covering both sub-cooled and saturated conditions. The reformulated Reynolds stress tensor removes non-physical sensitivities to porosity field and grid resolution, ensuring consistent hydraulic redistribution. The up-scaled viscosity correlation, derived from neptune_cfd simulations via URANIE optimization, significantly reduces mass-flux prediction errors and reveals the dominant contribution of dispersive mechanisms under the examined conditions. The revised diffusion and dispersion thermal diffusivity improve the modeling of inter-channel energy transfer and the prediction of onset of boiling in hotter channel, while the regime-dependent void-dispersion model correctly reproduced redistribution peaks and their evolution under high void fraction conditions. Together, these models provide a physically grounded representation of momentum and energy diffusion within the porous framework of CATHARE3 3D module, enhancing its predictive capability for single- and twophase turbulent mixing phenomena in a simple but representative geometrical configuration. These developments currently function as research prototypes, but the framework establishes a robust basis for future applications to more complex geometries.
Conventional exergy analyses of Nuclear Power Plants (NPPs) inherently depend on thermodynamic boundary selection, yet its influence on the interpretation and distribution of exergy destruction has not been systematically examined. This study proposes a Dual-Boundary Framework (DBF) for the thermodynamic and exergetic assessment of a conceptual Pressurized Water Reactor (PWR), implemented using the TESPy nonlinear network solver coupled with CoolProp. The model is validated against reference-PWR performance trends before evaluating two complementary exergy boundaries under constant thermal power of 3714.50 MW. The Fission-Based Boundary (Approach 1) employs the theoretical fission pseudo-temperature as the exergy source, yielding an exergetic efficiency equal to the thermal efficiency (35.80%) and total plant exergy destruction of 2344.42 MW, encompassing both intrinsic nuclear degradation and component-level irreversibilities. In contrast, the Average Fuel Temperature-Based Boundary (Approach 2) adopts representative average fuel pellet temperature (Tfp,avg = 800°C) as the finite-temperature heat source, excluding fission-related exergy destruction from the thermodynamic boundary. This yields an exergy input of 2699.82 MW, an exergetic efficiency of 49.25%, and a Carnot efficiency of 72.68%. An ambient-temperature (T0) sensitivity analysis and a criterion-filtered local sensitivity analysis over PSG = 6.8−8.0 MPa, corresponding to approximately Tfp,avg = 799−806°C, demonstrate complementary diagnostic responses of the two exergy boundaries. Approach 1 quantifies overall degradation of nuclear energy into useful work, whereas Approach 2 highlights engineering-domain power-conversion irreversibilities while revealing environmental and fuel-temperature effects on thermodynamic performance. The proposed DBF distinguishes intrinsic nuclear irreversibilities from potentially actionable engineering losses, providing a physically consistent framework for exergy assessment and thermodynamic evaluation of solid-fuel NPPs.
Annular fuel demonstrates significant application potential in marine nuclear power systems due to its advantages of high power density and lightweight design. However, ocean conditions can aggravate the unevenness of coolant flow distribution, which may lead to fluctuations in thermal-hydraulic parameters due to external disturbances such as platform rolling and thereby compromise reactor operational safety. This study experimentally investigates the flow distribution characteristics of a 5×5 annular fuel channel under steady-state and rolling motion conditions. Particle Image Velocimetry (PIV) was employed to quantify the transient flow response. The results indicate that the rolling-induced additional inertial force drives oscillations in the flow distribution ratio (φ), defined as the ratio of the inner-channel flow rate to the total inlet flow rate, at twice the rolling frequency. The oscillation amplitude of φ increases with the maximum rolling angle (θm) and decreases with the rolling period (T), exhibiting a distinct response lag relative to the rolling motion. Spectral analysis further reveals the presence of a higher-order harmonic component at 6fr, indicating nonlinear temporal response of the system to the periodic forcing. In contrast, the time-averaged flow distribution remains nearly unchanged under rolling motion within the investigated conditions and governed by the inherent differential flow resistance between the inner and outer channels. The rolling-induced flow redistribution is identified as a nonlinear forced response dominated by inertial effects. The nonlinear behavior results from the distortion of the periodic inertial forcing response and becomes more pronounced under stronger inertial forcing conditions.
The thermal-hydraulic characteristics of liquid lead-bismuth eutectic (LBE) are crucial to the reliability of advanced reactors used in emerging fields such as ocean exploration. To explore the relevant mechanisms of LBE natural circulation in ocean environments, the scaling analysis under ship rolling cases was conducted. The similarity criteria of scaled-down loops were derived. The prototype and the scaled model were simulated using the Fluent code to validate the similarity criteria, and the corresponding effects of rolling amplitudes and periods were analyzed. The circulation features of the prototype under different rolling cases can be accurately simulated, and the maximum deviation of the time-averaged parameters is less than 9.08%. The models with different length ratios all reproduce the dynamic trends of the mass flow rate and temperature difference of LBE in the prototype under different rolling conditions. The results show that the scaling method is applicable to the analysis of the LBE natural circulation system under rolling conditions, and is valuable for the design of scaled-down facilities under ship conditions.
The passive containment cooling system (PCS) plays an important role in limiting containment pressure and temperature during loss of coolant accidents (LOCA). Existing system codes are efficient for plant scale transient analysis but provide limited spatial resolution, whereas fully resolved computational fluid dynamics simulations of the containment and all PCS heat exchanger tubes are computationally prohibitive. To bridge this gap, PecTAF is developed as an integrated multiscale thermal hydraulic code that couples a one-dimensional primary side model with a three-dimensional porous medium model for the PCS heat exchanger secondary side and containment. The key condensation and coupling models are validated against the Pusan National University 10-rod bundle experiments with air, yielding an average relative error of 10.6% and a maximum deviation of 22.3% in total heat transfer power. PecTAF is then applied to a full scale PWR containment during the long-term cooling stage of LOCA. The results show that operation of the PCS heat exchangers induces large scale circulation under the combined effects of jet momentum and buoyancy. Within 3000 s, the mean steam partial pressure and temperature increase by 38% and 17 K without PCS operation, whereas the corresponding increases are limited to approximately 4% and 1 K with PCS operation.
To support dose assessment of routine carbon-14 (14C) releases from pressurized water reactors (PWRs), this study develops a design-stage integrated framework linking 14C production, chemical speciation, and release pathways. First, 14C production from the dominant 17O(n, α)14C and 14N(n,p)14C reactions is calculated using multigroup neutron fluxes and microscopic cross sections, and converted to an effective source term injected into an exchangeable carbon pool through an effective mobilizable fraction. Second, a five-compartment mass-balance model is formulated to describe interconversion, capture, and releases among aqueous-phase dissolved inorganic/organic carbon (DIC/DOC), gaseous 14CO2 and hydrocarbons, and accumulation on spent ion-exchange resins in the primary circuit. Monte Carlo uncertainty propagation provides design quantiles for pathway-specific releases and speciation fractions. Results show a stable fate hierarchy: gaseous releases dominate, resin solid waste is secondary, and liquid effluents are minimal. Speciation is strongly pathway dependent, with hydrocarbons dominating the gaseous releases, whereas inorganic DIC tends to dominate at the liquid discharge point due to speciation shift during liquid-waste treatment and storage. These results highlight the need for speciation-resolved monitoring for source term accounting and dose assessment. The framework produces a traceable design-envelope source term without plant-specific operating history and provides quantitative guidance for monitoring priorities, gaseous release management, and resin waste-load evaluation.
The depth of a repository for high-level radioactive waste is an important aspect with respect to both long-term and operational safety. In contrast to the minimum depth, no relevant constraints are given in the German legal regulations for a maximum value. To define such a limit, various safety-related aspects of claystone host rock are considered. This evaluation takes place within an early stage of the German site selection process. The available data are evaluated, and the respective assessments are discussed, ranging from aspects of construction and operation of the repository to its post-closure period. Furthermore, tunneling experience in relevant claystones is evaluated. It is concluded that a favorable depth is neither the maximum depth of existing underground excavations nor the maximum achievable with available techniques. Instead, it is the depth for which experience indicates that under the current assumptions, the extended operation of a nuclear waste repository can be well conducted without jeopardizing operational and long-term safety.The major argument for limiting the repository depth is robustness, which means that the repository as a whole can be operated safely, requires no extraordinary measures like resources, techniques, or time, and can be built with well-established techniques. It can be summarized that those requirements are unlikely to be met for a repository in claystone at depths greater than 1000 m.
Fast neutron flux in the reactor pressure vessel (RPV) is a critical parameter for assessing structural integrity. Reactor lifetime extension and power uprates increase the potential for exceeding fluence limits in the RPV extended beltline region. This paper investigates the sensitivity of fast neutron flux calculations in the RPV extended beltline region to numerical schemes and physical input parameters, with emphasis on spatial meshing, angular quadrature, source representation, fuel burnup, core-plate composition, cavity gap width, and concrete composition. The results indicate that, although stable results can be obtained in the traditional beltline region, the maximum deviation between 1.2 cm × 1.2 cm and 1.0 cm × 1.0 cm radial mesh sizes in the extended beltline region reaches 40.41% at an axial height of 644.53 cm. Relative to S16, the maximum deviation of S8 reaches 42.32%, indicating that the extended region requires higher angular discretization accuracy. The power-distribution analysis shows that pin-by-pin power representation of peripheral assemblies is necessary for reliable response calculations. Physical-parameter sensitivity analysis confirms that fuel burnup, core-plate composition, cavity size, and concrete properties can substantially influence flux predictions. These findings provide practical guidance for establishing conservative and traceable flux-evaluation models for extended beltline applications.
Chloride-based molten salt reactors are promising advanced nuclear systems because of their favorable neutronic characteristics and high actinide solubility, but their deployment is limited by severe corrosion of structural materials in chemically aggressive and irradiation relevant environments. This review aims to provide an integrated understanding of corrosion and mitigation in chloride-based molten salt reactors by connecting degradation mechanisms with broader materials challenges and design strategies. The analysis shows that corrosion is governed by the coupled effects of impurities, fission products, and irradiation rather than by a single factor. Moisture, oxygen, dissolved metallic ions, and other impurities destabilize salt chemistry, disturb redox balance, and accelerate selective dissolution and intergranular corrosion. Fission products further intensify degradation through distinct pathways: tellurium directly promotes embrittlement and grain boundary damage, whereas europium, antimony, and selenium contribute through redox perturbation or direct reaction layer formation. Irradiation adds another level of complexity by generating defects, elemental redistribution, phase instability, and gas accumulation, thereby altering both the material and its response to molten salt attack. These findings indicate that effective mitigation requires an integrated strategy combining salt purification and redox control with protective coatings, surface engineering, and alloy design. Overall, corrosion in chloride-based molten salt reactors should be understood as a coupled chemical and microstructural degradation problem, and reliable deployment will depend on structural materials and protection strategies validated under relevant reactors combined environments.
The free-piston Stirling generator (FPSG) provides a promising energy conversion solution for space nuclear power systems (SNPS) and deep space exploration. However, intense multiphysics coupling creates an extremely narrow steady-state operable domain. To ensure high-fidelity performance prediction, a third-order transient thermodynamic solver, previously validated against experimental benchmarks, is employed to capture nonlinear gas-solid heat transfer and fluid inertia effects. To overcome the high-dimensional optimization bottleneck, this study proposes a physics-informed active learning multi-objective optimization (AL-MOO) framework. This mechanism enforces strict thermodynamic conservation constraints and eliminates non-physical pseudo-optima frequently encountered in conventional data-driven methods. Macroscopic performance limits within strict safety constraints are precisely determined: maximum power output reaches 158.6 W (14.4% efficiency), and maximum efficiency peaks at 15.0% (114.9 W). Detailed energy flow breakdowns and entropy generation analysis reveal the underlying microscopic dissipation mechanisms. Pursuing maximum power inevitably triggers exponentially growing unsteady viscous hysteresis penalties, severe local thermal non-equilibrium, and regenerator enthalpy leakage. It simultaneously sacrifices impedance matching, causing significant electromagnetic reactive power losses. Monte Carlo analysis further confirms that the optimal design maintains robust performance under small manufacturing tolerances. This study quantifies the inherent competition between thermodynamic work capacity and energy efficiency, providing thermo-physical criteria for next-generation Stirling energy systems in nuclear applications.
Prestressed concrete containment vessel (PCCV) serves as the final barrier against the release of radioactive materials into the environment. Throughout the service life of a PCCV, time-dependent prestress losses happen due to the creep and shrinkage of concrete as well as the relaxation of tendons, which surely reduce the integrity of the PCCV and increase the failure probability correspondingly. Therefore, it is essential to investigate the fragility of the PCCV with different prestress loss levels under internal pressure. Firstly, a finite element model was established to simulate the experiment of a 1:4 scaled PCCV under increasing internal pressure at Sandia National Laboratories. Subsequently, the probability density evolution method (PDEM) was adopted for the probabilistic analysis of random structural responses, where material properties were assumed as the source of uncertainty. The ultimate state of the PCCV was evaluated using the global average strain, whose values of 0.4% and 1.5% corresponded to the functional failure and structural failure, respectively. Finally, a parametric sensitivity analysis was performed on the displacement response of the PCCV under internal pressure. The fragility of the PCCV under different prestress loss levels was also discussed. The results indicate that the Young's modulus of the prestressing tendons and the concrete have significant influences on the PCCV response. As the level of prestress loss increases, the integrity of the PCCV decreases. When the prestress loss level rose from 10% to 50%, the functional limit pressure capacity decreased by approximately 19%, while the structural limit pressure capacity decreased by about 7%.
The long-term immobilization of high-risk Cs-, Co-, and Ni-bearing radioactive waste remains a critical challenge due to the chemical instability and structural vulnerability of conventional borosilicate glasses under high waste-loading conditions. In this study, a series of potassium-alumino-borosilicate glass (BSKA) was systematically modified by substituting 25 mol% of B2O3 with Cs2O (BSKA:Cs), CoO (BSKA:Co), or NiO (BSKA:Ni), which serve as non-radioactive surrogates for Cs-, Co-, and Ni-containing radioactive wastes. The resulting glasses were comprehensively characterized using XRD, SEM/EDX, FTIR, DSC/TGA, Makishima-Mackenzie elastic modeling, density and porosity measurements, and leaching tests. XRD confirmed amorphous structures with peak shifts from 25.49° for BSKA base glass to 28.07°, 28.23°, and 27.73° for Cs-, Co-, and Ni-doped glasses, respectively. SEM revealed particle morphologies ranging from aggregated, rough surfaces (Cs) to anisotropic elongated particles (Ni). FTIR analysis indicated Cs-induced depolymerization, Co-induced densification, and Ni-induced moderate bond strengthening. Thermal analysis showed Tg values of 465 °C (BSKA:Cs), 452 °C (BSKA:Co), and 457 °C (BSKA:Ni), while Tc peaked at 609 °C for BSKA:Co, demonstrating enhanced crystallization resistance. Elastic moduli followed the trend BSKA:Ni > BSKA:Co > BSKA:Cs, with BSKA:Ni achieving the highest Young's modulus (Em), bulk (Km) modulus, and hardness (H). Leaching experiments confirmed low cumulative release (<0.35%) after 30 days, with transition Co- and Ni-doped glasses exhibiting superior hydrolytic resistance. These results highlight that targeted incorporation of Cs+, Co2+, and Ni2+ ions enables strategic structural tuning, optimizing mechanical strength, thermal stability, and chemical durability. Importantly, the preserved amorphous stability, overall structural homogeneity, high crystallization resistance, and sustained low leaching collectively demonstrate the strong suitability of these glasses for the stabilization of high-level radioactive waste under demanding conditions. This study provides a robust framework for designing next-generation borosilicate glasses capable of safely immobilizing high-level radioactive waste.
During severe accidents in a nuclear power plant, fission products released from the reactor core form aerosols in the containment. Aerosols are removed mainly through natural deposition and spray methods. This study applies the PISAA code to simulate these removal processes and validates the simulation results against the AHMED, TOSQAN and FABCS experiments. The simulation results agree well with the experimental data, with most predictions falling within ±20% error bands. Sensitivity analyses show that for natural deposition, larger particle size and higher density accelerate settling due to the stronger gravitational driving force. For hygroscopic aerosols, relative humidity has a measurable effect on the deposition rate, though its influence is weaker than that of particle size and density. For spray removal, higher flow rates and larger spray angles increase removal efficiency, whereas larger initial droplet diameters reduce it. The gains are nonlinear because high flow rates shorten droplet residence time and wide angles cause wall-directed losses. Smaller droplets improve removal by providing a higher number concentration and larger total surface area. The number of aerosol size sections also influences the predicted deposition rate. This study is of significance for improving the predictive reliability of PISAA for aerosol behavior under severe accident conditions.
Given the urgent global push toward a low-carbon and sustainable energy transition, liquid-fueled molten salt reactors (MSRs) are receiving renewed attention as a Generation IV nuclear technology capable of supporting long-term clean energy supply. MSRs exploit molten salts as both fuel and coolant, enabling a thorium-uranium breeding fuel cycle, high-temperature thermal efficiency, and intrinsic safety advantages. However, the liquid-fuel nature introduces unique safety analysis challenges. This review provides a comprehensive overview of representative MSR designs with emphasis on safety features and deterministic safety analysis requirements such as unique initiating events, thermophysical salt data, and modelling approaches for heat transfer, delayed neutron precursor transport, and salt solidification. The analysis reveals major gaps in current safety assessment practices, particularly uncertainties in salt properties and inadequate treatment of salt spill and off-gas scenarios. The review outlines future priorities including high-precision salt property measurements, tailored heat transfer correlations for complex geometries, efficient multiphysics simulation platforms, and new models for salt spill and post-spill, fission-gas transport, and natural circulation behaviors. The work aims to establish a consolidated technical foundation and roadmap to support future MSR design, safety validation, licensing, and deployment within sustainable energy systems.
Reprocessing of MOX fuel is a critical technology for sustainable nuclear energy. Its primary objectives include the extraction of reusable uranium and plutonium resources, and recovering long-lived actinide elements to provide conditions for transmutation disposal and eliminate long-term radioactive risks brought by nuclear energy development. The aqueous reprocessing technology has formed a relatively mature process technology in the reprocessing of ordinary thermal reactor spent fuel, so it is a practical route for the MOX fuel reprocessing. However, due to the significantly higher burnup, plutonium content, and fission product content of spent MOX fuel compared to thermal reactor spent fuel, there are substantial differences in the processes of shearing and dissolution, uranium/plutonium separation, as well as in radiation protection, criticality safety, and waste management. Consequently, the development of reprocessing technologies for spent MOX fuel is both highly significant and challenging. Currently, several countries such as France, Russia, India, and China, have initiated research on reprocessing technologies for high-burnup MOX fuel. This paper reviews the global research and practical experiments in spent MOX fuel reprocessing, with a focus on the chemical separation processes, and also provides a brief discussion on shearing and dissolution, backend processes, and waste disposal. As MOX fuel is the most mature fuel type for fast reactors, the current status of aqueous reprocessing technology for MOX fuel is mainly introduced.
