The thermal cycle sizing of a nuclear submarine’s propulsion loop requires accounting for a wide range of cruising velocities and transient phases, in both nominal and accidental operating modes. This loop comprises a single-phase heat exchanger on the reactor side and a condenser on the propulsion side, whose detailed temperature, velocity and pressure fields are computationally intensive to resolve over the full range of operating conditions. To drastically reduce the computational time, reduced order models (ROMs) have been built to calculate temperature, velocity and pressure fields in a simplified single-phase heat exchanger and in a small condenser. Three non-intrusive ROMs based on Proper Orthogonal Decomposition have been investigated. First, the global basis method constructs a single global reduced basis from high-fidelity training data and predicts a new solution by interpolating only the associated reduced coefficients for a new parameter (such as the inlet temperature of the inner tubes or the steam flow rate). Secondly, the Bi-CITSGM method (Oulghelou and Allery 2020) (Bi-Calibrated Interpolation on Tangent Subspace of Grassmann Manifold) follows a different interpolation strategy by interpolating not only the reduced coefficients, but also the spatial reduced basis for a new parameter. In contrast to these interpolation-based approaches, a physics-based reduced-order modeling strategy, referred to as the identification method, has been investigated. This approach assumes the form of the ROM and identifies its governing coefficients by solving an optimization problem. Once this ROM is constructed, it is solved for a new parameter to predict the corresponding system response. Unlike the two previous methods, which are purely data-driven, this method is physics-based because it relies on the identification and resolution of a low-dimensional system of differential equations that describes the physical behavior of the process. The results obtained with these methods have shown a satisfactory agreement with results obtained with the high fidelity solvers. In addition, a drastic reduction of the computational time has been highlighted.
Monoclinic zirconia has been shown to play a significant role in enhancing the corrosion resistance of zirconium alloy cladding materials. However, it has also been demonstrated that monoclinic zirconia affects the thermal transport properties of the cladding material. Simultaneously, the irradiation process induces the formation of vacancy defects, including isolated vacancies and vacancy clusters, which further deteriorate the thermal properties. Nevertheless, the quantitative influence of vacancy concentration and defect clustering on the thermal conductivity of monoclinic zirconia under reactor-relevant conditions remains unclear. In this paper, non-equilibrium molecular dynamics (NEMD) simulations are used to obtain the thermal conductivity of monoclinic zirconia containing vacancy defects and to systematically account for finite-size effects. The temperature dependence of the thermal conductivity exhibits good agreement with available experimental data over the investigated range. The inhibitory effect of vacancy defects on thermal conduction is interpreted in terms of enhanced phonon scattering. These results reveal that, at a fixed total vacancy concentration, the aggregation of individual vacancies into clusters can attenuate this scattering effect and increase the effective thermal conductivity compared with isolated vacancies. This study provides the fundamental insight into heat transport in irradiated monoclinic zirconia and offers the desirable parameters in order of magnitude for assessing oxide-layer thermal conductivity and thermal margins in pressurized water reactors under normal operation.
The cost of primary magnesium production is highly dominated by the energy required to produce the metal. This, along with the current geographical concentration of its production, makes the magnesium price vulnerable as energy-production paradigms begin to rapidly change. In this work, four nuclear-integration pathways for the efficient supply of heat and electricity to magnesium production from a nuclear power plant are considered. A production route using electrified metallothermic reduction is compared to a nuclear-heat-integrated electrolytic pathway. It is shown that the latter can achieve significantly lower primary energy consumption, 261 MJ/kg Mg compared with 416 MJ/kg Mg. The electrolytic pathway can also reduce emissions down to around 2.1 kgCO2/kg Mg, compared to around 34 kgCO2/kg Mg in the Pidgeon process. Beyond energy and emissions, the four pathways considered are also compared to their ability to reduce locational dependence of the production plant and reduce process complexity. This work also highlights future advances in magnesium production with a potential scheme for the efficient extraction of magnesium from seawater with coproduction of desalinated water and the potential of demand-side response from magnesium electrolysis. This is the first work to consider the full integration of primary magnesium production using nuclear cogeneration, and it is hoped the study paves the way for the analysis of economic nuclear integration within metal extraction and manufacturing.
In pressurized water reactors (PWRs), the LanceHead sealing joint—a critical interface of the in-core instrumentation system's thimble located on the reactor pressure vessel—serves as a key component in maintaining the integrity of the primary circuit pressure boundary. Its long-term sealing reliability is directly crucial to nuclear reactor safety. The sealing quality of these joints is primarily controlled through tightening torque, with the precise torque-preload relationship being central to ensuring sealing performance. To advance prediction and control capabilities for joint sealing performance, this study developed an autonomous online leakage-tightening test rig operating at 0–155 bar and temperatures from ambient to 303 °C. Systematic experiments revealed variation patterns of leak-tight torque with temperature and pressure: under conventional operating conditions, torque exhibits positive correlation with both parameters, showing a significant jump beyond 100 °C due to medium vaporization; under extreme high-temperature/high-pressure conditions, torque demonstrates fluctuating characteristics. Leveraging experimental data, we employed the NRBO-XGBoost (Newton-Raphson-Based Optimizer with eXtreme Gradient Boosting) algorithm to establish a leak-tight torque prediction model. This model demonstrated significantly superior accuracy compared to BP neural networks, support vector machines, and other benchmarks, achieving an R2 of 0.99 with a maximum prediction error of 5.7 N·m. Concurrently, preload measurement experiments were conducted to derive an empirical formula correlating axial preload with tightening torque. Furthermore, a preload prediction model was developed using the NRBO-XGBoost algorithm, demonstrating significantly superior performance (R2 = 0.99, RMSE = 106 N) over conventional empirical formulations. These results comprehensively validate the effectiveness, reliability, and engineering practicality of the proposed method for high-precision prediction of sealing joint parameters. This research provides theoretical foundations and technical support for optimizing sealing performance and safety assurance in nuclear-grade joints.
To address the demand for large-scale production of the medical isotope molybdenum-99 (Mo-99), this study evaluates two neutron-activation production schemes using natural molybdenum in a CANDU reactor: a large-channel moderator-interlayer target and a composite fuel-bundle target. Neutronic performance was assessed using the Reactor Monte Carlo code (RMC) and compared with a small-channel reference target. The large-channel target is straightforward to implement and, with heavy-water interlayers, effectively mitigates spatial self-shielding, increasing the total yield by approximately 4320% compared with the small-channel target scheme, albeit with a 17.0% reduction in specific activity. The composite fuel-bundle scheme, in which the central fuel element is replaced by a natural‑molybdenum target rod, increases the Mo-99 production density by 55.0% and provides a relative total-yield increase of about 12.5 times, while introducing only a small full-core power perturbation (0.95%). Simplified reactivity and heat-source estimates further indicate negligible influence on overall reactor power production under the limited replacement conditions considered. Overall, the composite fuel-bundle design shows strong potential for scalable activation-based Mo-99 production in CANDU reactors.
During decommissioning, there is a requirement to reduce the physical size of the range of components that are used within nuclear facilities, allowing for their safe recycling or disposal as waste. Laser and plasma cutting systems are routinely used for size reduction tasks within the nuclear decommissioning sector, and within the nuclear industry, are often robotically deployed to remove workers from radioactive environments. However, issues arise where cutting operations fail to completely sever a joint or materials, rendering further challenges related to item movement and storage. This requirement to ensure that cutting operations are fully complete drives research into in-process monitoring that can provide data on the quality of the cut, the veracity, and total severance of structures being decommissioned, such as gloveboxes.This paper introduces a robotically deployed air-coupled ultrasound in-process monitoring system that inspects cut severance during plasma cutting of 3 mm-thick 316 stainless steel plates representative of nuclear glovebox material.In this work, a plasma cutting robot is equipped with non-contact ultrasonic transducers and associated electronics to generate and detect an A0 Lamb wave mode in these metals and provide cut quality monitoring and screening.Experiments demonstrated that fully cut and uncut plate segments, as well as small uncut sections as narrow as 10 mm wide can be successfully detected. The results also show that Lamb waves remain detectable when inspecting cuts that do not fully penetrate the material.
Pressurized heavy water reactors (PHWRs) fueled with natural uranium (NU) are constrained by low discharge burnup (∼7500 MWd/MTHM), high spent fuel volumes, and a positive coolant void reactivity (CVR) that poses a safety challenge under loss-of-coolant accident (LOCA) conditions. Clean Core Thorium Energy, Inc. (CCTE) is developing Advanced Nuclear Energy for Enriched Life (ANEEL), a thoria-urania (ThO₂-UO₂) mixed oxide fuel using high-assay low-enriched uranium (HALEU) and a burnable neutron absorber, designed as a “plug and play” replacement for existing PHWR NU fuel bundles. This paper presents physics, thermal-hydraulic, and fuel performance analyses supporting the technical feasibility of ANEEL fuel in a CANDU-6 reactor.Fuel composition optimization using DRAGON/DONJON yielded a design achieving an average discharge burnup of over 44,000 MWd/MTHM — approximately six times that of NU fuel — while reducing the full-core CVR from 16.4 mk to below 13.3 mk for an equilibrium core. Core-follow simulations confirm that maximum channel and bundle powers remain within licensing limits throughout the fresh-to-equilibrium transition. Independent MCNP calculations confirm that reactivity device worths deviate by no more than 16% from NU values and that the fuel and moderator temperature coefficients are more negative for ANEEL compared to NU fuel. CATHENA LOCA simulations show peak sheath and centerline temperatures for ANEEL are lower than for NU fuel. PEGASUS finite element analysis confirms structural integrity throughout irradiation, with fission gas pressure remaining below coolant pressure and end-of-life cladding strain of much less than 1%. These results demonstrate that ANEEL fuel achieves substantially higher burnup with improved safety margins relative to NU fuel, with manageable impacts on existing PHWR core design and control systems.
Ballooning and burst separate-effects tests are widely used to reproduce cladding behavior under loss-of-coolant accident (LOCA) conditions and to validate transient fuel-performance models. A high-temperature burst rig was developed at the HUN-REN Centre for Energy Research to perform isothermal tests up to 1100 °C under controlled linear pressurization rates (1 kPa/s to 1 MPa/s) in inert atmosphere. To explore fuel relocation and potential dispersal pathways at burst, selected tests used simplified fragmented-pellet simulators assembled from precision ceramic grinding balls with two representative size fractions. High-speed imaging (up to 62,500 fps) captured the rapid ejection of fragments through the burst opening.After testing, the geometry and wall-thickness distribution of ballooned and burst specimens were quantified using three complementary techniques: (i) metallography-based thickness measurements from stitched optical micrographs, (ii) X-ray computed tomography (CT) with surface extraction to point clouds followed by automated binning into axial slices and angular sectors, and (iii) 3D optical profilometry of both outer-surface and inner-surface scans enabled by a custom automated actuator, using negatives produced by flexible replica casting. The techniques are complementary and generally consistent in overlapping regions, and together they reveal pronounced asymmetry of deformation and localized wall thinning before the burst. The resulting experimental and metrology workflow provides code-facing metrics (e.g., minimum remaining wall thickness and its axial/circumferential location) to support improved burst criteria and uncertainty quantification in fuel-performance modelling during LOCA.
To clarify the influence of xenon (Xe) addition and its molar fraction on the performance of supercritical carbon dioxide (sCO2) compressors and cycles, this study investigates CO2-Xe mixtures using a combined approach of numerical simulation and thermodynamic analysis. An sCO2 centrifugal compressor and an idealized regenerative Brayton cycle are modeled to systematically examine the effects of Xe addition on component-level and system-level performance.The results show that: (1) Xe addition shifts the compressor efficiency map toward higher mass flow rates and improves high-flow off-design performance. However, high Xe fractions may reduce the low-flow operating margin because the slope of the pressure-ratio–flow-coefficient curve decreases near the low-flow side. (2) Xe addition effectively suppresses the equilibrium phase-transition tendency within the impeller by reducing low-temperature and low-pressure regions. It also promotes a more uniform flow field and mitigates internal losses under high-flow conditions. (3) At the cycle level, Xe addition shows the potential to improve the idealized regenerative Brayton-cycle performance by reducing the total exergy destruction rate. The main contribution comes from a substantial decrease in recuperator exergy destruction rate, even though compressor irreversibility increases.Overall, the results indicate that CO2-Xe mixtures have potential to improve the high-flow operating characteristics of sCO2 compressors and the idealized thermodynamic performance of CO2-based mixture Brayton cycles. These findings provide guidance for the preliminary design and optimization of compressors in advanced nuclear power conversion systems.
Liquid metal heat pipes are being evaluated for a variety of space and microreactor applications. The operating envelope of these devices is dominated by the capillary limit, which is determined largely by the capillary radius and permeability of the heat pipe wick structure. To simultaneously enhance both parameters, wrapped mesh screen wicks that leave a gap between the mesh and cladding have been proposed. However, characterization of as-manufactured annular wick samples remains limited. This work used rate-of-rise testing to measure the capillary radius and permeability of rectangular and annular wick samples made from 316 stainless steel 100-mesh screen. The annular samples, around 21 mm in outer diameter, were then measured with cladding that created an annular gap of around 0.63 mm. Two possibilities exist when calculating the capillary radius and permeability from rate-of-rise data with an annular gap: using the bare mesh cross-sectional area and porosity or using the total mesh plus gap area with an averaged porosity. Both methods were evaluated, and a comparison to a simple analytic model suggested that the latter was more representative of the actual phenomena. However, this resulted in only a 1.3 to 1.6-times increase in the wick figure-of-merit compared to a three-times increase using the former method employed by other researchers. These differences need to be further investigated by comparing the rate-of-rise method with independent measurements of permeability and capillary radius.
This technical note focuses on the corrosion response of austenitic 316L stainless steel in boron-free small modular reactor (SMR) environments containing KOH and NH3. The oxide scale analysis showed a duplex oxide layer, with an outer layer rich in Fe and an inner layer rich in Cr, in both KOH and NH3-SMR environments. Electrochemical methods, which included Electrochemical Impedance Spectroscopy and Mott-Schottky analysis, were employed to examine the stability of the oxide layers. The post-exposure electrochemical findings indicated the formation of a marginally protective oxide layer in the KOH-SMR environment. The mechanism of oxide formation in KOH and NH3-SMR environments was also discussed.
In this paper, the mechanical shim (MSHIM) control strategy is implemented in the Hualong Pressurized Reactor (HPR1000) under base load operation. To obtain a lower adjustment frequency of boron concentration and stable axial power distribution, the design of MSHIM control strategy is executed in the HPR1000. The reactivity control-banks with different overlap fractions, initial lead bank positions, and AO-bank positions are separately inserted into the HPR1000 to investigate the core neutronics characteristics. The high-fidelity deterministic neutronics program CRANE, which uses the method of characteristics to solve the transport equation, is used to obtain the core neutronics results during operation. An automated research method for determining the operational time of the control bank position based on CRANE is developed to quickly achieve the MSHIM base load operation of the core. The results show that the HPR1000 core with 20% overlap fraction of reactivity control-banks, 0 steps of initial lead bank position, and 195–210 steps of R-bank position has a smaller adjustment frequency of boron concentration and more stable power distribution, which is recommended for the HPR1000 core during MSHIM base load operation. This work provides a valuable reference for the design and application of MSHIM control strategy in the HPR1000.
After the occurrence of cracks in the main coolant circuits of pressurized water reactors (PWR), resulting in long and expensive plant outages, the thermal mixing of hot and cold water streams in tees was investigated with the help of dedicated experiments and sophisticated simulations. In 2023, OECD/NEA initiated the international “Thermal Mixing and Fatigue in a T-Junction with a Dead Leg Benchmark”. It is based on Vattenfall experiments dealing with the flow and heat transfer phenomena in a T-junction with a horizontal main branch and vertical blind-end branch. While the water in the latter is stagnant and cold, the one in the main branch is hot and has high velocity. The Vattenfall experiments have high reactor safety relevance, since they mimic the conditions in the safety injection and residual heat removal system lines (blind-end branches) that are connected to the main reactor coolant legs. This paper presents the computations, carried out at GRS, for the open test case of this OECD/NEA benchmark. The comparison between calculated and measured temperatures in the fluid and solid domains of the tee geometry shows that the hot fluid penetration depth as well as the temperature oscillations are qualitatively and quantitatively well predicted by the Unsteady Reynolds-Averaged Navier-Stokes (URANS) approach implemented in ANSYS CFX.
Thermal fatigue caused by turbulent thermal mixing can compromise the integrity of solid components through crack initiation, potentially leading to failure. Understanding the flow and heat transport physics is essential for predicting and preventing such degradation. This study presents a Direct Numerical Simulation (DNS) of thermal mixing in a circular pipe T-junction with conjugate heat transfer. A hot branch flow is injected orthogonally into a cold main flow, with equal inlet flow rates and a bulk Reynolds number of 5300. The simulation includes solid walls of thickness 0.1 pipe diameter and employs two passive temperature scalars with Prandtl numbers 1 and 0.025, with Neumann boundary condition at the exterior solid boundaries. The passive scalar treatment of the temperature fields limits the scope of the present study to momentum-driven thermal striping under negligible buoyancy conditions. Compared with the DNS studies reported in the literature, the principal novelties of the present work are the application of a realistic three-dimensional T-junction geometry incorporating conjugate heat transfer through the solid walls, together with the consideration of both low- and unity-Prandtl number fluids. The primary objective of the present study is to provide fundamental insight into the flow and heat-transfer phenomena introduced by these novel features, thereby extending the current understanding beyond that available from existing DNS studies in the literature, and to generate high-fidelity data for assessment and modelling of thermal fatigue in nuclear reactor piping. Particular focus is placed on quantifying wall temperature fluctuations, their penetration into the pipe wall, and the influence of fluid Prandtl number on thermal loading relevant to structural integrity analyses. The results reveal key flow structures, including deflection of the branch flow, recirculation zones, and turbulence generation primarily at the shear layer. Temperature fluctuations are highest at the interface of the deflected branch and main flows, with unity-Prandtl flows showing stronger fluid fluctuations and low-Prandtl flows inducing larger solid wall fluctuations. Turbulent heat fluxes and power spectral density analyses further characterize the transport and spectral content of temperature fluctuations.
Conventional high-load helium compressors are prone to significant tip leakage and rotor dynamics issues. The rim-driven motor configuration effectively mitigates these challenges; however, it introduces a tip leakage cavity that alters the flow characteristics at the rim tip region of the high-load helium compressor. To analyze the influence of the leakage cavity geometry on the compressor performance and to provide practical design guidelines for optimization, numerical simulations were conducted to investigate the effects of radial clearance width, axial clearance width, and different sealing configurations. The results show that reducing the axial clearance from 1.5 mm to 0.5 mm increases the peak efficiency by 4.7%, while reducing the radial clearance over the same range yields a 3.7% improvement. Furthermore, the optimized sealing configuration (Case F) improves the peak efficiency point of high-load helium compressor by 5.7% without modifying the clearance widths.
Critical heat flux (CHF) is a key boundary parameter for the thermal-margin assessment of nuclear power systems. The Groeneveld 2006 look-up table (LUT) offers broad coverage but, as an empirical engineering correlation, exhibits systematic bias under extreme geometries, while purely data-driven models can lose credibility outside their training envelope. This study develops a LUT-guided grey-box framework in which the LUT acts as an empirical engineering prior and a multilayer perceptron learns its discrepancy with experiments, through an additive-residual (GB-Add) and a multiplicative-correction (GB-Mult) architecture trained with a weighted penalty loss. The models are built on 3501 experimental points for uniformly heated round tubes covering P∈[0.1,20]MPa, G∈[21,7961] kg/m2/s, X∈[−0.49,0.91] and D∈[1.02,37.47]mm, and are evaluated in two stages: a random 80:20 split, and leave-one-source-out and parameter-range holdout tests that probe generalisation and bounded extrapolation. On the random split GB-Mult is competitive (R2=0.9664, MAPE =9.89%, ±15% in-band ratio 76.75%), although a random-forest baseline attains the highest in-band ratio (80.74%) and the lowest MAE. The grey-box value lies in an interpretable, LUT-anchored correction: the factor K stays near unity, the CHF–X monotonicity and non-monotonic pressure trend are inherited, and split-conformal prediction intervals reach 90.4%/95.6% coverage. The holdout tests delimit the scope, as the correction degrades under severe source shift and at extreme geometry. The framework is thus an interpretable, scope-limited template for uniformly heated round tubes.
