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Phebus is an experimental facility that represents a scaled-down version of the French 900 MWe pressurized water reactor (PWR) with a ratio of 1/5000. In order to study phenomena occurring during severe accidents in light water reactors, five experiments (FPT-0 to FPT-4) were performed under different fuel and cooling conditions. This paper provides a comprehensive description of the Phebus FPT-1 experiment, MELCOR and ASYST computer codes, discusses the modelling approaches employed and compares the calculated results with the experimental data. To quantitatively assess the results, the Figure of Merits of selected parameters were calculated for both computer codes using ACAP software.

The spatial difference of the thermal-hydraulic state in reactor core is easy to lead to local high temperature, so the risk of material burning and fuel melting is introduced under the safe conditions of average temperature. The fine computational fluid dynamics analysis of reactor core can carefully identify the extreme values and distribution of sub-millimeter fluid velocity, pressure, temperature and other parameters in reactor core. However, in traditional research, the states of flow resistance, thermal resistance, flow mixing and other states are measured by the macroscopic integration of lines, surfaces and volumes, and the fine data of tens of millions or hundreds of millions of computing grids are mostly converted into hundreds or thousands of sets of average data, the physical state is reduced from three-dimensional to two-dimensional or one-dimensional state, and the spatial resolution is reduced from submillimeter to decimeter or lower, and the fine data cannot be effectively and fully utilized. Therefore, the fine flow characteristics of reactor core are studied. The distribution of secondary flow intensity, temperature, and temperature inhomogeneity coefficients in the axial direction in the domain of a 15×15 rod bundle fuel assembly was first analyzed, and the flow characteristics induced by the spacer grid with the mixing vane were found to be periodic on each span domain. After that, the spontaneous configuration characteristics of the fluid were explored, and the fine flow analysis method based on the mechanical gear system was designed and named as flow gear system. The structure of the flow gear system in the fluid domain and the characteristics of energy storage, drag reduction, the rate of decline in energy quality, spontaneous optimization, maintenance of large flow mixing and flow identification were investigated, and the order of the turbulence of reactor core was clarified. In addition, the transverse flow character with fixed-wall conditions in a small domain like 5×5 rod bundle analyzed with flow gear system clarifies the limitations, which was unable to identify the large flow gear and long transverse flow structures, so that some thermal-hydraulic physical information was lost. The results show that the transverse flow analysis strategy based on the flow gear system is valuable to guide the research of the transverse flow in fuel assembly, and the important roles in the transverse flow keeping the quality of kinetic energy have been clarified, which can be seen as a new and useful strategy and methodology for the fine thermal-hydraulic data in reactor core.

Sodium-cooled fast reactors are considered the preferred reactor type for fourth generation nuclear energy systems due to their high thermal efficiency, high fuel utilization efficiency, inherent safety, and unique advantages of achieving a closed fuel cycle to reduce long-lived radioactive waste. They are also a hot research topic in the international nuclear energy field. An overview of the development history and future directions of sodium-cooled fast reactors and thermal-hydraulic fields on an international scale was provides, and the main problems currently faced in development of sodium-cooled fast reactor thermal-hydraulic fields were pointed out. Based on the current development status and challenges in China, an in-depth analysis of three key core aspects in the field of thermal-hydraulic research was conducted, including the flow and heat transfer characteristics of coolant sodium, core thermal-hydraulic analysis methods, and design and capability verification of removing core decay heat through natural circulation. The research achievements on an international scale was summarized, the technological challenges encountered and the future development trends of research were analyzed, and the current research status and future research directions in these fields in China were explored. These analyses and summaries aim to provide guidance for the technical challenges of improving safety and economy in China’s sodium-cooled fast reactor technology, and also provide important reference for the thermal-hydraulic design and optimization of future sodium-cooled fast reactors, in order to promote the further development of China’s sodium-cooled fast reactor technology.

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

A detailed computational fluid dynamics code named JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research (JUPITER) for the thermal-hydraulics around the beam window (BW) of accelerator-driven system (ADS) was used to confirm the validity of the thermal-hydraulics design tool based on the ANSYS Fluent. The Fluent uses the Reynolds-averaged Navier–Stokes (RANS) model and can quickly calculate the turbulent flow around the BW as a BW design tool. First, the results of JUPITER were compared with the experimental results using a mock-up BW system in water to confirm the validity of JUPITER. This study confirmed that the numerical results are in good agreement with the experimental results. Thus, JUPITER could be used as a benchmark code. A benchmark simulation for the Fluent calculation was also performed using validated JUPITER to demonstrate the applicability of JUPITER as an alternative for experiments. Therefore, the mean values around the BW agreed with each other (e.g. the mean velocity profile for the stream and horizontal directions). Therefore, results confirmed that JUPITER demonstrated a good performance in validating the thermal-hydraulics design tool as a fluid dynamics solver. Moreover, Fluent has sufficient accuracy as a thermal-hydraulics design tool for the ADS. [ABSTRACT FROM AUTHOR]

In the IFMIF-DONES neutron source a single accelerator line delivers 125 mA of deuterons at 40 MeV to a flowing liquid lithium target with a free surface facing the deuteron beam.Neutrons with fusion-relevant spectrum are generated by a stripping reaction and used for material samples irradiation.This report presents a numerical thermal–hydraulic and structural analysis using updated design of the target system and neutronics analysis results. Simulation scope includes the target assembly, vacuum chamber with beam ducts, outlet channel, quench tank vessel and connecting elements.Simulations have been conducted with the commercial CFD code Simcenter Star-CCM + . The computational domains consist total of 47.4x106 fluid and 14x106 structure cells.The following aspects have been analyzed: • Temperature distribution in lithium and target system structure during beam-on operation; • Transient mixing of lithium in the quench tank with temperature change during the beam switch-on phase. • Thermal displacements and stresses in the TA and QT structures.Simulations using the new distribution of nuclear heating show no significant changes in the temperature field in liquid lithium. Calculated unwanted temperature hotspots in the target structure demonstrate the need for optimization of the target design.

This paper introduces a multiphysics code, named nemoFoam, designed within the OpenFOAM environment to address the coupling of neutron and photon transport with thermal-hydraulics in nuclear reactor simulations. The code, conceived both for fusion and fission applications, employs a modular approach where the neutronic module is currently based on multi-group neutron diffusion equations, including a mono-kinetic treatment for photons. The thermal-hydraulic module is built on the standard OpenFOAM solver. The paper focuses on presenting the key features of nemoFoam and discussing the structure of the code, the nuclear and thermal-hydraulic models, and the coupling strategy between them.In order to assess the performance of the neutronic module, the code is applied to a fusion case study, indeed a benchmark against the Serpent Monte Carlo code for neutron and photon transport is performed, applying nemoFoam to the Affordable, Robust and Compact (ARC) fusion reactor design. Simulation results demonstrate good agreement with Monte Carlo benchmarks, emphasizing the potential of the code for multiphysics simulations. The modular design of nemoFoam allows users to extend the implemented model equations to study additional phenomena, focus only on selected aspects independently and, potentially, to add new solvers in each module.

AbstractComparison of results for global responses predicted by different multiphysics simulations of benchmark problems may fail to reveal potentially significant local modeling issues. An examination of code interactions in coupled simulations can provide more information, which may help identify potential modeling issues that went unnoticed during verification (and validation) of the individual codes, or may call into question approximations otherwise deemed reasonable at the individual code level.We illustrate this challenge for the case of coupled neutronics/thermal-hydraulic transient simulations using one of the problems and contributed results documented in International Atomic Energy Agency TECDOC-1994. This recently published report documents the specifications of four numerical multiphysics pressurized heavy water reactor (PHWR) transient challenge problems and the results contributed by 10 participants. Our work is based on the pump rundown problem, where TECDOC-1994 suggests that differences in modeling and methods employed in thermal-hydraulics may be the dominant factor in the observed differences. We performed a more detailed assessment with two different multiphysics coupled computational frameworks using NESTLE-C/ARIANT and PUMA/RELAP-5. We also studied a pump seizure transient, a more challenging variant of the pump rundown transient. Several aspects were investigated: comparisons of standalone results, sensitivity to gap modeling, selection of boundary conditions at the pressurizer, and an examination of correlations used in ARIANT and RELAP-5.Our assessment goes beyond the results for global parameters and dives into details of predictions at the channel level. This paper briefly describes the PHWR pump rundown transient problem and a pump seizure variant, the computational methods employed, and the areas investigated, and discusses some selected results. [ABSTRACT FROM AUTHOR]

Blanket is of vital importance for engineering application of the fusion reactor. Nuclear heat deposition in materials is the main heat source in blanket structure. In this paper, the three-dimensional method for thermal-hydraulics/neutronics coupling analysis is developed and applied for the full-scale module of the helium-cooled ceramic breeder tritium breeding blanket (HCCB TBB) designed for China Fusion Engineering Test Reactor (CFETR). The explicit coupling scheme is used to support data transfer for coupling analysis based on cell-to-cell mapping method. The coupling algorithm is realized by the user-defined function compiled in Fluent. The three-dimensional model is established, and then the coupling analysis is performed using the paralleled Coupling Analysis of Thermal-hydraulics and Neutronics Interface Code (CATNIC). The results reveal the relatively small influence of the coupling analysis compared to the traditional method using the radial fitting function of internal heat source. However, the coupling analysis method is quite important considering the nonuniform distribution of the neutron wall loading (NWL) along the poloidal direction. Finally, the structure optimization of the blanket is carried out using the coupling method to satisfy the thermal requirement of all materials. The nonlinear effect between thermal-hydraulics and neutronics is found during the blanket structure optimization, and the tritium production performance is slightly reduced after optimization. Such an adverse effect should be thoroughly evaluated in the future work.

Lead-bismuth reactor has the advantages of high inherent safety, strong natural circulation capacity, long operation life and rapid transient response, and has a good application prospect in special applications such as ocean engineering. The Marine device will be affected by wind and waves in the ocean and produce rolling motion, which has a series of effects on the thermal and hydraulic characteristics of the system. In order to study the thermal-hydraulic characteristics of lead-bismuth loop system under rolling motion, the size and structure of the loop system were determined, the CFD geometric model was established, the calculation mesh was drawn and the calculation model of the natural circulation characteristics under rolling motion was developed. Based on the established rolling motion calculation method, the numerical simulation study of rolling motion was carried out under the design conditions, and the influence of rolling motion on the natural circulation capacity and heat transfer characteristics of the loop system was studied. Afterwards, experiments were carried out to compare the experimental data with the simulated data to verify the correctness of the model and the reliability of the conclusions. The relative deviation between simulated and experimental fluid temperature is generally less than 5.3%, and the average relative deviation is less than 4.2%. The relative deviation of wall temperature is generally below 2.1%, and the average relative deviation is 1.67%. In addition, the maximum deviation between simulation and experiment of the rolling motion flow is within 3%. The comparison results demonstrate that there is a little gap between the simulation results and the experimental results, and the gap is within the acceptable range, which indicates that the calculation results are reliable. The simulation results show that the rolling motion will cause periodic fluctuations in the natural circulation flow of the loop system. The period of the flow fluctuation is consistent with the motion period, and the amplitude of the fluctuation increases with the maximum rolling motion and the decrease of the motion period. When the maximum rolling angle is 20° and the rolling period is 7.5 s, the instantaneous fluctuation of the flow rate reaches a maximum of 140%, and the reverse flow phenomenon occurs at some moments. The rolling motion will also cause periodic instantaneous fluctuations in the flow and heat transfer characteristic parameters of the system. The law of this fluctuation is similar to the fluctuation law of natural circulation flow. The maximum instantaneous fluctuation of Nu number reaches 515%, and the maximum time average volatility can reach 66%. The results obtained in this study have important engineering application value and scientific significance for the design and optimization of lead-bismuth loop system under rolling motion.

The nuclear safety assessment involving large transient simulations is forcing the community to develop methods for coupling thermal-hydraulics and neutronic codes and three-dimensional (3D) Computational Fluid Dynamics (CFD) codes. In this paper a set of dynamic boundary conditions are implemented in OpenFOAMⓇ in order to apply zero-dimensional (0D) approaches coupling with 3D thermal-hydraulic simulation in a single framework. This boundary conditions are applied to model pipelines, tanks, pumps, and heat exchangers. On a first stage, four tests are perform in order to assess the implementations. The results are compared with experimental data, full 3D CFD, and system code simulations, finding a general good agreement. The semi-implicit implementation nature of these boundary conditions has shown robustness and accuracy for large time steps. Finally, an application case, consisting of a simplified open pool with a cooling external circuit is solved to remark the capability of the tool to simulate thermal hydraulic systems commonly found in nuclear installations.

BackgroundThe enrichment of nuclear fuel may significantly influence the fuel performance of a reactor.PurposeThis study aims to explore the effect of U3Si2-Al plate-type fuel enrichment on its performance,MethodsThe fuel performance analysis code BEEs-Plates, neutronics Monte Carlo code OpenMC, and the one-dimensional system analysis code ZEBRA were coupled together within the MOOSE (Multiphysics Object-Oriented Simulation Environment) framework. Then the coupling code was employed to conduct the multiphysics coupling calculation for JRR-3 (Japanese Research Reactor No.3) fuel assembly enriched at 15%, 20%, and 25%. The data exchange among the three codes were realized with the help of interpolation transfer method implemented in MOOSE. Additionally, neutron physical parameters and fuel performance parameters after 231 d of operation were analyzed when the average fuel consumption of the module was 125.71 GWd∙tU-1.ResultsThe calculation results indicate that the max power density of fuel assembly enriched at 25% is 18% higher than the assembly enriched at 15%. Due to the high thermal conductivity of aluminum, the temperature difference in the fuel assembly is almost negligible whilst there is a significant difference in the fast neutron fluence. The results of fuel temperature and fast neutron fluence show that the volumetric strain is more obviously affected by the fuel temperature. Specifically, the plastic strain of the assembly with 25% enrichment is approximately 40% higher than that of the assembly with 15% enrichment.ConclusionThe analysis results of this study suggest that the assembly with a higher enrichment is more prone to failure.

The investigation of thermal-hydraulic parameters during steam-line break (SLB) accidents is performed by applying the personal computer transient analyzer (PCTRAN) simulator model of the VVER-1200 nuclear power plant (NPP). Five cases, namely, 0.005 m2 (Case-1), 0.01 m2 break (Case-2), 0.02 m2 break (Case-3), 0.04 m2 (Case-4), and 0.08 m2 (case-5) of SLB accident inside containment with the concurrent loss of AC power have been simulated. There was no variation in the timing of the trip of the reactor coolant pumps, the main feedwater pumps, or the turbine in any of the five SLB accidents. However, the reactor scram's onset time varies slightly between the five scenarios. Pressure and temperature in the reactor coolant system (RCS) quickly reached a peak following the start of the SLB accident, fell shortly after the reactor scram, and eventually stabilized in all cases. In comparison to the larger breaks in the SLB accident, the smaller breaks result in a higher RCS temperature and pressure. After the SLB accident, the pressurizer's liquid level rises and then quickly drops in all cases. The break mass flow rate from the steam line rapidly increases until the occurrence of the reactor scram and then decreases to a stabilized value. Steam generator A has a faster rate of heat removal rate than steam generator B, and its pressure and liquid level decrease more quickly than those of steam generator B. The thermal power of the reactor, peak cladding temperature, and fuel temperature showed a rapid drop after the initiation of the SLB accident. There was no increase in these parameters from the initial state of the simulation. The radiation in the air of the reactor building and steam line was very low during the simulation period. Therefore, there was no violation of the safety aspects of the SLB accident of the PCTRAN simulation of the VVER-1200 NPP model.

Several conductor designs featuring High Temperature Superconducting (HTS) stacked tapes for fusion coils are being proposed. These conductors are planned to operate in time-varying magnetic field and current; thus, the estimation of AC losses is fundamental for the conductor design and the accurate analysis of its performance in operation. The case study of an HTS conductor proposed for the hybrid (HTS-LTS) Central Solenoid coil for the EU DEMO tokamak is considered in this work. Here, a numerical model based on the finite element method (FEM) and the H-formulation is used, in order to estimate the hysteresis losses. The FEM model is first benchmarked against available analytical formulae as well as available literature data. Then it is applied to the real case operational scenario. It is shown that for the conductor design analyzed, the coupling losses are orders of magnitude lower than the hysteresis ones. The impact of the hysteresis+coupling losses on the temperature margin of the coil is assessed with a thermal-hydraulic model. It is shown that the heat generated in the HTS layers is partially transferred to the LTS layers, leading these layers to quench. An alternative conductor concept is also analyzed, showing that, however, in the top and bottom modules of the CS coil, due to the bending of the magnetic field, a too large heat deposition is present.

In the design of magnetic confinement nuclear fusion power plants, the breeding blanket (BB) plays a crucial role since it must fulfil key functions such as tritium breeding, radiation-shielding, and removal of the heat power generated by the plasma. The latter task is achieved by the first wall (FW) and breeding zone (BZ) cooling systems, which in the water-cooled lithium–lead (WCLL) BB employ pressurized water. Different arrangements of BZ coolant conduits have been investigated in the recent past to identify an efficient layout, which could meet the structural materials’ operational temperature constraint and which could provide the optimal coolant outlet temperature. However, most of the computational fluid dynamic (CFD) analyses that have been carried out until now have been focused on the equatorial WCLL elementary cell of the central outboard segment (COB). The aim of this work is to broaden the analysis to other relevant locations in the blanket. An assessment of the design of the cooling system of the COB bottom-cap elementary BZ cell has been identified as a top design priority due to its different geometry and thermal loads. The cooling efficiency of the BZ and FW systems is investigated to assess if the coolant-appropriate design conditions are matched and the temperature distribution in the cell is analyzed to identify the onset of hot spots. Different layouts of the FW systems are proposed and compared in terms of thermal–hydraulic reliability.

The LVR-15 research reactor is operated with a tube type fuel IRT-4M. Due to fuel’s unique concentric square annular shape with, coolant flow is subject to significant pressure driven crossflow. Detailed calculation of such flow patterns is beyond capabilities of standard system codes used for the thermal hydraulic safety analysis. To assess safety of core designs consisting of the IRT-4M fuel assemblies, a new subchannel code is under development at Research Centre Rež and Czech Technical University in Prague. Computer program SUBSALS is single phase, steady state, subchannel solver in development. Rigid computational mesh describes IRT-like geometries, and it is also suitable for other specific doubly connected ducts. In this contribution, general description of calculation procedure and computational capabilities is presented along with code-to-code comparison. It could be used as proof of concept for future improvements in thermal hydraulic calculations of research reactors.

The helical-coiled once-through steam generator (H-OTSG) is widely used in small modular reactors due to its compactness and higher heat transfer efficiency. In this study, an experimental and numerical simulation study of the thermalhydraulics characteristics of HOTSG under different thermal power conditions was carried out, and the experimental results were used to verify the onedimensional system code SGTH1D. The configuration consists of 85 helically coiled tubes divided into ten layers according to the coil diameter. Both steadystate experiments and flow instability experiments were conducted under thermal power condition ranging from 06 MW to 23 MW. Firstly, the steadystate experimental results show the difference between the shellside inlet temperature and the tubeside outlet temperature increases with thermal power, which is due to the decreasing of the heat transfer capacity of the HOTSG. Sensitivity analysis of the system parameters shows that the shellside pressure has little effect on the average heat transfer coefficient. And the average heat transfer coefficient also is insensitive to tubeside pressure and shellside inlet temperature under low thermal power conditions. However, under low thermal power conditions, the average heat transfer coefficient increases with the decrease of the tubeside pressure and the increases of the shellside inlet temperature. Secondly, the flow instability experimental results show that the increase of the thermal power can stabilize the HOTSG and decrease the inlet throttling of the flow instability thresholds because the length of the subcooled singlephase region increases with the thermal power. The sensitivity analysis of the system parameters shows that the increase of the tubeside pressure and the decrease of the shellside inlet temperature are beneficial to the stability of the system. The influence of the shellside parameters on the flow instability threshold is weaker than that of the tubeside parameters. Finally, a numerical study on the thermal-hydraulics characteristics of the configuration was carried out in the present study. By comparing the experimental data with the numerical simulation results of the SGTH1D code, the accuracy of the SGTH1D code in predicting the steadystate heat transfer characteristics and flow instability of configuration was verified. The SGTH1D code can accurately predict the heat transfer rate of the HOTSG under various thermal power conditions, and the prediction errors of shellside and tubeside outlet temperatures are within ±1 ℃. The numerical results of flow instability characteristics of the SGTH1D code are conservative at low thermal power condition, and the error of numerical results is within ±20% when the thermal power is larger than 12 MW.

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Natural circulation in a two-phase water loop, with imposed wall power, is investigated at atmospheric pressure both experimentally and numerically, using the system-scale code CATHARE. Low-pressure loops are prone to instability due to strong differences of density between the liquid and the vapor. Particular attention is devoted to analyzing the boundaries of the flow reversal regime with respect to flow cross section aperture (related to friction forces and experimentally performed using regulation valves) upstream and downstream of the heated test-section. At low heat flux, the flow is stationary, whereas at higher heat flux, flow reversal appears. The system-scale code is amenable to reproduce the flow reversal regime boundaries explored experimentally. An analytical criterion is used to highlight that boiling may become unstable and flow reversals appear. Results show that the stable flow region boundaries can be extended by increasing the upstream pressure loss coefficient (valve closing) in order to reach large exit void fractions α ≈ 1. In configurations where the upstream pressure losses dominate, stable boiling points with high mass flux and high heat flux are observed. On the opposite, larger downstream pressure loss coefficients strongly favor instability and flow reversal. [Display omitted] • The system-scale code CATHARE reproduces satisfactorily the flow reversal regime boundaries. • At low pressure, the instabilities appear simultaneously with boiling. • Increase of upstream (resp. downstream) pressure losses improves (resp. decreases) flow stability. [ABSTRACT FROM AUTHOR]

• Benchmark study of NACIE-UP facility with uniform and non-uniform power distribution is studied by ANSYS Fluent. • Heat transfer characteristics of NACIE-UP at three different steady states have been analyzed and the conjugate heat transfer is advised to be taken into consideration. • Verification and Validation (V&V) analysis is performed for stand-alone CFD code to compare with code-to-code results. • RANS SST k-w model and Cheng&Tak correlation are proved suitable to wire-wrapped bundle. The NACIE-UP (NAtural Circulation Experiment-Upgrade) facility is located at the ENEA Brasimone Research Centre (Italy) and is designed to investigate the thermal–hydraulic properties of Lead-Bismuth Eutectic (LBE) within a wire-spaced assembly under uniform and non-uniform power distribution conditions. This benchmark exercise proposed by ENEA is based on previous experiment results, intending for development and validation of CFD/TH system code, stand-alone CFD simulation, etc. In this study, the thermal–hydraulic characteristics of Fuel Pin Simulator (FPS) are analyzed, which contains inlet bend and outlet section. ANSYS Fluent solver is applied with the COUPLE algorithm of RANS SST k - ω model. Cheng&Tak correlation and computational conditions are implanted into Fluent in the form of UDFs. Moreover, calculations were performed for ADP10, ADP06 and ADP07. The results show that the RMSEs of ADP10SS1 and ADP10SS2 are 2.52 K and 3.69 K, 7.05 K and 8.48 K for ADP06SS1 and ADP06SS2, respectively, which indicate that the adopted models have good prediction ability for the temperature of the experimental conditions. Meanwhile, the heat transfer characteristics of different sub-channels and sections are investigated and compared with the correlations of Ushakov, Mikityuk, Kazimi and Carelli. The experimental and simulation values of the uniformly heated ADP10 are in good agreement with the correlations, while there is a large deviation for the non-uniformly heated ADP06. Therefore, it is necessary to develop new correlations for the heat transfer characteristics of LBE in wire-wrapped bundle. Above all, this study provides quantitative and qualitative analysis for verification and validation of CFD/TH system code and stand-alone CFD simulation. [ABSTRACT FROM AUTHOR]

Small Modular Reactors (SMRs) have gained significant attention as the next generation of nuclear reactors due to their unique features, such as improved economic viability and safety. This study focuses on the SMART reactor, the first certified design of an SMR, and investigates various Thermo-Neutronics aspects using an integrated coupling scheme of nuclear codes. The Neutronics cell and core calculations are performed using the DRAGON/DONJON codes, while the COBRA code is employed for thermal-hydraulic calculations. The analysis considers parameters including axial and radial power peaking factors (PPFs), critical heat flux (CHF), minimum departure from nucleate boiling ratio (MDNBR), axial average coolant temperature in hot channel and core, and maximum fuel temperature. These parameters are evaluated using the developed coupling system. Code-to-code verification is conducted to ensure the accuracy of the results, demonstrating good agreement. The coupling approach reveals significant effects compared to stand-alone code modeling, emphasizing the importance of considering the coupling between Neutronics and thermal-hydraulics, especially during the cycle length. This study contributes to the understanding of SMRs, particularly the SMART reactor, by providing insights into key parameters and their interdependencies. The integrated coupling scheme enhances the accuracy of the analysis and highlights the significance of considering coupling effects in the modeling process during the steady state and the cycle length. • The main LW-SMR core Neutronics and TH parameters have been investigated in standalone and coupling modes. • The proposed coupling N-TH approach reveals significant effects compared to stand-alone code modeling. • Current research contributes valuable insights into SMR main core parameters and their interdependencies. • DRAGON/DONJON and COBRA codes have been utilized for Neutronics and TH calculations. [ABSTRACT FROM AUTHOR]

The control rod withdrawal (CRW) transient is one of the most important safety issues for the early design phase of sodium fast reactors (SFR). CRW transients can cause core melting, which can lead to severe accidents. To ensure the integrity of the core, it is crucial to study CRW transients in detail since early design phase of the reactor. Two approaches are used in numerical modeling of transients at CEA. Reference codes (APOLLO3 (Schneider et al., 2016), GERMINAL (Lainet et al., 2019) for example) are used to perform detailed and precise calculations, recently these tools are being coupled to form multiphysics platforms such as described in Pascal et al. (2021); Li et al. (2023). Despite being robust and validated, this type of simulation is not adapted to design phases due to the need in computing resources it requires. The second type of approach, selected in this paper, is to use simplified, parametric tools with a limited validity range. We present here a new numerical tool called MORPHEE, which is capable of simulating CRW transients. MORPHEE is a multiphysics and multiscale tool that includes neutronics, thermal-hydraulics, and fuel behavior. The tool has been developed to provide a comprehensive understanding of the CRW transient and to assess the safety of SFR. It is based on MACARENa (Droin et al. 2020) and OCARINa (Herbreteau et al. 2018) tools. They have been validated for simulation of thermal-hydraulics and fuel pin behavior, respectively. MORPHEE uses finite-differences methods to solve thermal-hydraulics equations, neutron kinetics equations and mechanical equations. A machine learning model (Jeannin et al., 2022) is also used to predict fuel cladding heat transfer through the gap. The simulations have demonstrated the importance of considering the multiphysics and multiscale aspects of the CRW transient to predict the behavior of the reactor during an accident. In conclusion, MORPHEE is a numerical tool that can simulate the CRW transient in SFR. It can be used to study the safety of SFR during the early design phase and to optimize the design to ensure the safety of the reactor. • Multiphysics methodology for Sodium Fast Reactors transients simulation. • Control Rod Withdrawal (CRW) phenomenological-based modeling with MORPHEE tool. • CRW modeling tool to guide design of SFR during the conception phase regarding safety. • Fuel cladding heat exchange coefficient and impact on transients over power. • Neutronics simplified simulation of flux shape deformation. [ABSTRACT FROM AUTHOR]

• Sapienza University of Rome developed a coupled approach to simulate FFTF ULOF transient. • TH model was refined to consider inter-assembly heat transfer and RTD inertia. • This allows a better representation of row 6 PIOTA and loop temperatures. • A complete NK/TH coupled model was developed by using PHISICS/RELAP5-3D® and ECCO codes. • The coupled codes results showed a good agreement with the experimental data. A Coordinated Research Project (CRP) on the benchmark analysis of Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) has been held under the sponsorship of the International Atomic Energy Agency. The CRP aims to improve understanding of loss of flow events in fast reactors, as well as to assess capabilities of existing computer codes against experimental data. FFTF is a 400 MWth Sodium-cooled Fast Reactor (SFR) owned by the U.S. Department of Energy, operated from 1980 to 1992. The CRP involves the LOFWOS Test #13, belonging to a series of unprotected transients performed as part of the Passive Safety Testing program. In this framework, the Nuclear Engineering Research Group of Sapienza University of Rome has developed an integrated multiphysics modelling based on a coupled approach with RELAP5-3D© (thermal-hydraulics) and PHISICS (neutron-kinetics) codes. The present paper discusses the results obtained by the authors after the participation in the benchmark open phase. As suggested by the organizers, a two-step methodology was followed. Firstly, calculations were run by imposing the reactor power as boundary condition and focusing on the FFTF thermal–hydraulic behaviour. A good accordance was detected between the numerical results and the experimental data, above all for what concerns the reference core outlet temperatures. Then, the simulations were repeated by using a coupled approach. Calculation outcomes demonstrated the capability of the selected code suite (PHISICS/RELAP5-3D®) to reproduce the reactor transient behavior and its suitability to be used as an effective numerical tool to simulate liquid metal fast reactors accidental scenarios where thermal–hydraulic and neutron-kinetic phenomena are strongly coupled. [ABSTRACT FROM AUTHOR]

After a hypothetical loss of primary coolant accident (LOCA) in a pressurized water reactor (PWR), the reactor safety systems inject water into the core to cool it. During this re-flooding process, a flow of steam and dispersed droplets (DFFB — Dispersed Flow Film Boiling) is formed downstream of the quench front. This two-phase flow serves to cool the section of the nuclear fuel rods that are not yet in contact with water. On the other hand, depending on the severity of the accident, the rods may deform, causing the reduction in the passage area of the coolant fluid. In addition, these deformations cause the steam flow (formed in the quench front) to be partially deviated to less deformed sub-channels, a phenomenon known as steam flow redistribution. Because of this, the cooling capacity of these partially blocked sub-channels may be compromised. During the present work, different tests were performed with the COLIBRI experimental bench in order to evaluate the influence of steam deviation on the cooling of a partially blocked sub-channel. The COLIBRI experimental bench allows the evaluation of the heat flux extracted by a typical steam and droplet flow of a LOCA in different blockage configurations. In this study two blockage configurations were considered: an intact configuration (0%) and a severely blocked configuration (90%). Given that steam convection is presumed to be the primary heat transfer mechanism in DFFB, we aim to assess the impact of reduced vapor mass flow rates in blocked sub-channels resulting from vapor redistribution on internal two-phase flow heat dissipation. The results show that steam deviation negatively impacts the heat flux extracted by the internal steam and droplet flow (heat flux could degrade by up to 43%). However, it was observed that the decrease in heat flux between the steam and the wall could be partially compensated by the droplets. In addition, the influence of the residual power on the heat flux extracted by the DFFB was also investigated, but no significant effect of this parameter on the heat flux value was revealed. • PDA and IRT tests to evaluate blockage ratios and droplet flow rates on sub-channel cooling. • Steam deviation degrades heat transfer in severely blocked subchannel two-phase flow. • Droplets present in dispersed flow film boiling enhance heat transfer. • Droplets partially compensate the heat dissipation reduction due to steam deviation. • Residual power levels are investigated to understand their effect on heat flux in a closer-to-realistic LOCA scenario. [ABSTRACT FROM AUTHOR]

The overall performance of steam generators plays a significant role in ensuring the safety of a nuclear power plant (NPP) operation. The analysis of thermal-hydraulic parameters during a steam generator tube rupture (SGTR) event of VVER-1200 NPP is conducted by applying the personal computer transient analyzer (PCTRAN) simulator. Four cases, namely, 25%, 50%, 75% and 100% of one tube rupture in two steam generators with the concurrent loss of AC power have been performed. Among the four cases, major variation in time was not observed for the occurrence of the reactor scram, reactor coolant pump trip, main feed-water pump trip, and turbine trip. The pressure and the temperature of the reactor coolant system (RCS) increase rapidly to a peak value due to event initiation, and drop promptly after the reactor scram. The stabilized pressure and temperature of the RCS are higher for the smaller break size of the SGTR. The secondary pressure of the steam generator is also increased to a peak value, followed by an increasing and decreasing trend, in turn, due to the repeated opening and closing of safety relief valves of the steam generators. The liquid level of the pressurizer is increased rapidly due to the liquid surge towards the pressurizer after the event and it is stabilized after the opening of the safety relief valve. The stabilized liquid level of the pressurizer and the steam generator is higher for the smaller break size of the SGTR. The earlier emergency core coolant injection to the reactor was required for the larger break size of the SGTR. There is no increase in the peak cladding temperature and the peak fuel temperature during the calculation period for all these cases. The results of this study provide a valuable understanding of SGTR events with the concurrent loss of AC power for the PCTRAN model of the VVER-1200 NPP.

The analysis of pressure changes in water and steam flow along the entire once-through steam boiler tube system at the 650 MWe lignite fired Unit was carried out after the long operational period of almost 30 years without a chemical cleaning of the pipe system. A significant increase of the hydraulic resistance was recorded in comparison to the design values in the evaporator tubes, both in the water heating zone and in the evaporation zone with the two-phase mixture flow. The cause of this adverse effect is the formation of a wavy surface of the magnetite layer deposited on the inner wall of tubes. The significantly increased absolute roughness on the inner wall of tubes was determined on the basis of measured pressure drops at sections of the tube system and numerical calculation performed with a thermal-hydraulic model of single-phase and two-phase water and steam tube flows. The model is based on the steam-water two-fluid model with mass, momentum and energy balance equations and mechanistic correlations for the steam-water interface transfer processes. The model is suitable for the single and two-phase flow parameters prediction in the span from subcritical to supercritical and near critical pressure conditions. Conducted calculations provide detailed insight in the fluid pressure, velocity and temperature change in the boiler tube system. After the long operational period the pressure drop in evaporating tubes almost doubles in comparison to the design values. The presented results are a support to boiler's design, operational procedures and maintenance.

In the design of magnetic confinement nuclear fusion power plants, the breeding blanket (BB) plays a crucial role since it must fulfil key functions such as tritium breeding, radiation-shielding, and removal of the heat power generated by the plasma. The latter task is achieved by the first wall (FW) and breeding zone (BZ) cooling systems, which in the water-cooled lithium–lead (WCLL) BB employ pressurized water. Different arrangements of BZ coolant conduits have been investigated in the recent past to identify an efficient layout, which could meet the structural materials' operational temperature constraint and which could provide the optimal coolant outlet temperature. However, most of the computational fluid dynamic (CFD) analyses that have been carried out until now have been focused on the equatorial WCLL elementary cell of the central outboard segment (COB). The aim of this work is to broaden the analysis to other relevant locations in the blanket. An assessment of the design of the cooling system of the COB bottom-cap elementary BZ cell has been identified as a top design priority due to its different geometry and thermal loads. The cooling efficiency of the BZ and FW systems is investigated to assess if the coolant-appropriate design conditions are matched and the temperature distribution in the cell is analyzed to identify the onset of hot spots. Different layouts of the FW systems are proposed and compared in terms of thermal–hydraulic reliability. [ABSTRACT FROM AUTHOR]

The LVR-15 research reactor is operated with a tube type fuel IRT-4M. Due to fuel's unique concentric square annular shape with, coolant flow is subject to significant pressure driven crossflow. Detailed calculation of such flow patterns is beyond capabilities of standard system codes used for the thermal hydraulic safety analysis. To assess safety of core designs consisting of the IRT-4M fuel assemblies, a new subchannel code is under development at Research Centre Rež and Czech Technical University in Prague. Computer program SUBSALS is single phase, steady state, subchannel solver in development. Rigid computational mesh describes IRT-like geometries, and it is also suitable for other specific doubly connected ducts. In this contribution, general description of calculation procedure and computational capabilities is presented along with code-to-code comparison. It could be used as proof of concept for future improvements in thermal hydraulic calculations of research reactors. [ABSTRACT FROM AUTHOR]

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Bulky metallic structures are needed in the toroidal field (TF) superconducting magnets for fusion applications to withstand the large Lorentz forces acting on the winding. The pulsed coil operation during a plasma scenario and the fast current discharge, or a plasma disruption event in off-normal operating conditions, cause transient magnetic fields, inducing eddy currents in the TF structures. The eddy currents generate heat in the structures heating in turn the winding pack, eroding the temperature margin: such power deposition is a key input for thermal-hydraulic (TH) analyses. However, the computation of eddy currents in fusion magnets is a challenging topic since a transient, fully 3D electromagnetic (EM) model is required. The EM problem is solved here by means of the finite element (FE) open source code FreeFEM++. First, the correct implementation of the EM problem is verified by means of suitable benchmarks against both simple analytical cases and the results obtained with state-of-the-art FE commercial codes on the DTT TF coil, used as a reference geometry. Then the EM code is applied to the evaluation of the magnetic fields and eddy currents induced in the same reference coil during the normal (static) and off-normal (transient) operation; the output of the EM analysis is used as input to the TH analysis carried out with the 4C code, aimed at computing the temperature margin evolution during the transient.

Natural circulation technology and helical-coiled tube steam generator are widely adopted in marine nuclear power system design due to the inherent safety feature of natural circulation and high heat transfer coefficient of helicalcoiled steam generator. The unique structural characteristics of helical-coiled pipe causes some special phenomena such as secondary flow effect inside the tube, and the heat transfer characteristics are more complicated than those in circular pipe. Under the environment of ocean motion, flow patterns and heat transfer characteristics of the fluid inside the helical tube will also change significantly under the influence of the nonlinear additional force. Thus relevant research is necessary to clarify the influence of ocean motion to the flow and heat transfer characteristics inside helical tube. Based on this background, a natural circulation system experiment loop under ocean motion was built, and the research on the natural circulation characteristics of the system and heat transfer characteristics of helical tube under ocean conditions was conducted in this paper. Firstly, experiment of single phase flow and heat transfer characteristics inside the helical tube under a variety of typical single and coupling ocean motions was conducted. The results show that driving force of natural circulation drops under inclination movement, and the steady flow rate and pressure drop are less than those of static condition, flow resistance coefficient under inclination condition is slightly lower than static condition. The influence of trim on the natural circulation capacity of the system is greater than that of heel. Average heat transfer characteristics inside the helical tube under ocean conditions were studied furtherly, it is found that the heat transfer capacity along the side near the central axis is significantly weaker than that along the side far from the central axis, which can be 24 times lower rolling motion, declines the average heat transfer coefficient of helical tube while pitching motion promotes it. An empirical relation of heat transfer coefficient in helical tube under ocean motion was established based on experimental data, and the error between the experimental value and predicted value is less than 10%. The influence of ocean motion on the occurrence point of natural circulation flow instability was studied, results show that inclination movement can make the occurrence of natural circulation flow instability ahead. Geometric model of the experimental loop was built employing a marine condition system analysis code based on RELAP5 code, and the transient simulation of rolling motion was carried out. Calculation results are in good agreement with the flow rate and pressure drop experiment data, which verifies the reliability of the program applied to natural circulation system analysis under ocean motion. Results of this study can offer reference for natural circulation analysis of floating and marine reactors and performance evaluation and design of helical-coiled tube steam generator under ocean condition, and also provide data support for the development of related analysis code.

Nuclear thermal propulsion is an enabling technology for future space missions, such as crew-operated Mars missions. Nuclear thermal propulsion technology provides a performance benefit over chemical propulsion systems by operating with light propellants (e.g., hydrogen) at elevated engine chamber conditions. Therefore, nuclear thermal propulsion reactor cores exhibit high propellant velocities and elevated propellant and fuel temperatures, subsequently leading to relatively high thermal stresses and geometrical deformation. This paper details the numerical approach to solve the thermo-elastic equations, which was implemented into the recently developed ntpThermo code. In addition, this paper demonstrates the extension of the Basilisk multiphysics framework to perform full-core coupled neutronic, thermal-hydraulic, and thermo-mechanical analysis of nuclear thermal propulsion reactors. The analyses demonstrate and quantify thermo-mechanical feedback, which for the investigated cases, acted to reduce maximum fuel temperatures and pressure drop across the fuel element channels. Thermo-mechanical feedback had a significant impact on the mass flow distribution within the reactor core and, thus, a substantial impact on solid-material temperatures and stresses, but only a minor impact on reactivity and local power distributions. Sensitivity studies revealed that the friction factor correlation applied to perform the analysis has a significant impact on the pressure drop across the fuel element channels. The most important observation of this research is the importance of incorporating the thermo-mechanical feedback within an integrated multiphysics solution sequence to enable the consistent design of future nuclear thermal propulsion systems. [ABSTRACT FROM AUTHOR]

At present, most of the widely used system codes for nuclear safety analysis are one-dimensional, which cannot effectively simulate the flow field of the reactor core or other structures. This is true even for the system codes containing three-dimensional modules with limited three-dimensional simulation function such as RELAP-3D. In contrast, the computational fluid dynamics (CFD) codes excel at providing a detailed three-dimensional flow field of the reactor core or other components; however, the computational domain is relatively small and results in the very high computing resource consuming. Therefore, the development of coupling codes, which can make comprehensive use of the advantages of system and CFD codes, has become a research focus.In this paper, a review focus on the researches of coupled CFD and thermal-hydraulic system codes was carried out, which summarized the method of coupling, the data transfer processing between CFD and system codes, and the verification and validation (V&V) of coupled codes. Furthermore, a series of problems associated with the coupling procedure have been identified, which provide the general direction for the development and V&V efforts of coupled codes.

When a nuclear reactor with rectangular fuel assemblies runs for a long time, impurities and debris may be taken into coolant channels, which may cause flow blockage, and the blocked fuel assemblies might be destroyed. Therefore, the purpose of this study is to perform a thermal-hydraulic analysis of a rectangular fuel assembly by STAR-CCM+, under the condition of one subchannel with 80% blockage ratio. A rectangular fuel assembly of the International Atomic Energy Agency (IAEA) 10 MW material test reactor (MTR) is chosen. In view of the gasket material taken into the coolant channel is close to the single side of the coolant channel, in the flow blockage accident of the Oak Ridge Research Reactor (ORRR), a new blockage category called single side blockage is attempted. The blockage positions include inlet, middle and outlet, and the blockage is set as a cuboid. It is found by simulations that the blockage redistributes the mass flow rate, and large vortices appear locally. The peak temperature of the cladding is maximum, when the blockage is located at the single side of the coolant channel inlet, and no boiling occurs in all blockage cases. Moreover, as the height of the blockage increases, the damage caused by the blockage increases slightly.

The molten salt reactors are a kind of nuclear reactor that had received additional attention worldwide due to their inherent safety features and the possibility of use the Thorium, a more abundant resource than Uranium, as a fuel. Despite of this promise concept and works being developed or under course, some aspects related to their thermal hydraulics should be better understand. In this work, the initial thermal-hydraulics results obtained using the ANSYS-CFXÒ code are presented. The simulations were conducted in 3D for 1/16 of a cylinder volume that represents the molten reactor core. The results show a good agreement for this simplified geometry and for a theoretical power density distribution between the benchmark work and those obtained within this for the temperature. An evaluation for the other properties and for different turbulence models, among others, are planned for future.

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect its stress conditions, mechanical behavior and thermal-hydraulic performance. A reliable numerical method is of great importance to reveal the complex evolution of mechanical deformation, flow redistribution and temperature field for the plate-type fuel assembly under non-uniform irradiation. This paper is the first part of a two-part study developing the numerical methodology for the thermal-fluid-structure coupling behaviors of plate-type fuel assembly under irradiation. In this paper, the thermal-fluid-structure coupling methodology has been developed for plate-type fuel assembly under non-uniform irradiation condition by exchanging thermal-hydraulic and mechanical deformation parameters between Finite Element Model (FEM) software and Computational Fluid Dynamic (CFD) software with Mesh-based parallel Code Coupling Interface (MpCCI), which has been validated with experimental results. Based on the established methodology, the effects of non-uniform irradiation and fluid were discussed, which demonstrated that the maximum mechanical deformation with irradiation was dozens of times larger than that without irradiation and the hydraulic load on fuel plates due to differential pressure played a dominant role in the mechanical deformation.

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect stress conditions, mechanical behaviors and thermal-hydraulic performance of the fuel assembly. This paper is the Part II work of a two-part study devoted to analyzing the complex unique mechanical deformation and thermal-hydraulic characteristics for the typical plate-type fuel assembly under irradiation effect, which is on the basis of developed and verified numerical thermal-fluid-structure coupling methodology under irradiation in Part I of this work. The mechanical deformation, thermal-hydraulic performance and Mises stress have been analyzed for the typical plate-type fuel assembly consisting of support plates under non-uniform irradiation. It was interesting to observe that: the plate-type fuel assembly including the fuel plates and support plates tended to bend towards the location with maximum fission rate; the hot spots in the fuel foil appeared at the location with maximum thickness increment; the maximum Mises stress of fuel foil was located at the adjacent location with the maximum plate thickness increment et al.

Nuclear power plants play a critical role in providing clean and sustainable energy. Ensuring the safety of these plants is of utmost importance to the nuclear industry. In this regards, thermal-hydraulic computer codes are essential for the simulation and understanding of the behavior during both normal and accidental conditions. In this context, LOCUST 1.2 is a newly developed thermal-hydraulic code by China General Nuclear, which aims at simulating the steady-state and accidental behavior of HPR-1000, a Pressurized Water Reactor design. This paper presents the validation of LOCUST 1.2 using the Advanced Nuclear Technologies assessment methodology from the Universitat Politècnica de Catalunya. The validation focuses on a nodalization created for the Large Scale Test Facility from the Japan Atomic Emergency Agency. In particular, four tests from the OECD/NEA ROSA 1 and 2 projects were selected. The selected tests encompass scenarios such as Anticipated Transient Without Scram, Intermediate Break Loss-Of-Coolant Accident, Steam Generator Tube Rupture, and Main Steam Line Breaks. The ANT-UPC methodology provides a comprehensive phenomenological assessment combining qualitative and quantitative analyses with the help of Best Estimate Plus Uncertainty calculations. All four tests are divided in phenomena through a Phenomena Identification and Ranking Tables (PIRT) to then perform the assessment of each phenomena individually, finally yielding an assessment matrix. Due to length constraints, this article only presents the detailed description of four key phenomena. However, the full PIRT table together with the assessment for each phenomenon is provided. The overall findings indicate that LOCUST 1.2 demonstrates a good capability to accurately reproduce most of the phenomenology observed in all four tests. • Validation of LOCUST 1.2 thermal-hydraulic code. • Use of the ANT Assessment Methodology for phenomenological assessment of LSTF tests. • Use of BEPU to determine the uncertainty of LOCUST 1.2 thermal-hydraulic code. [ABSTRACT FROM AUTHOR]

The demonstration of the safety, high performance and reliability of any man-made device relies heavily in experimental programs. For nuclear reactor safety demonstration, several experimental facilities are used for both testing the design and the validation of computational codes. In the field of thermal hydraulics, most experimental facilities are built at a smaller scale, thus the scaling analysis is a requirement in the licensing process of the nuclear power plant. This includes the analysis of the scaling capabilities of the computer codes, the scale and design differences with the licensed reactor and the methodologies applied for licensing. In collaboration with the Spanish regulatory Body, several Spanish university groups have contributed considerably to the development and application of scaling methodologies. This article provides an overview of the most relevant contributions presented in the last decades. The activities deal mainly with the use of system codes in scaling analysis. Specifically the works can be classified in the use of K v scaling analyses, analysis of counterpart experiments, plant application and the scaling of uncertainties. Finally, the future challenges that will be addressed in relation to scaling are highlighted with emphasis to the contribution to the safe operation of current and future nuclear power plants. • Spanish contributions to the development and application of scaling methods. • Advances and past applications of scaling analysis such as Kv scaling in Spain. • Future challenges regarding scaling methodologies in thermal hydraulics. [ABSTRACT FROM AUTHOR]

• System level LBE thermal-hydraulics analysis code is developed. • Coupling effect of LBE oxidation corrosion and thermal-hydraulics is considered. • Influence of oxidation layer on heat transfer characteristics of LBE reactor is conducted. For Lead-Bismuth Eutectic (LBE) alloy cooled reactor, oxidation corrosion is very important and would affect the thermal hydraulic characteristics of the reactor. In this paper, based on LETHAC code developed by Xi'an Jiaotong University (XJTU) for system level thermal hydraulic analysis LBE reactor, a simulation module for oxidation corrosion is developed and added. The influence of oxidation layer on the heat transfer characteristics of the reactor is conducted. Based on LESMOR reactor designed by XJTU, a simulation calculation is carried out under the long-term operation conditions of the reactor considering the growth of oxidation layer. The distribution of the oxide layer in the reactor, the change of temperature appreciation caused by the oxide layer over time and the distribution in the reactor are obtained. The presence of oxide layer will affect the fluid temperature and wall temperature of reactor. The longer the run time, the greater the impact. This paper could provide solid analysis methodology reference for the design of LBE reactor. [ABSTRACT FROM AUTHOR]

Liquid metals are employed as a coolant in liquid metal fast reactors (LMFR) and are considered breeders and coolants for future power-producing fusion reactors. The interaction of liquid metals with other coolants, such as air and water, is one of the possible occurrences in liquid metal-cooled reactors. Although the SIMMER-III computer code is currently in the validation and verification phase, it is a prospective candidate code for correctly investigating possible accidents. This study aims to determine the stability and well-posedness of the SIMMER-III code Eulerian-Eulerian two-fluid model (TFM) under any such accident scenario. This work also considers the influence of the virtual mass force and diffusion forces in momentum on TFM stability in accelerated liquid metal, steam, and non-condensable gaseous flows throughout all flow regimes. The characteristics method is used to assess the ill-posed nature of TFM for all types of accidents and accident scenarios in a hypothetical simplified system model with several components (Lead–Lithium, non-condensable gases, and water vapor). It has been discovered that the analysis findings vary from the air and water two-component two-phase flows (characteristics roots spectrum and error growth rate patterns) and are particularly sensitive to the diffusion and virtual mass coefficients. This is because liquid metals have a higher density than liquid water and steam, resulting in strong virtual mass forces and weak diffusion forces in liquid-metal and gas two-phase flows. Because of the extensive range of possible interactions between fluxes of different components, producing an accurate representation of diffusion in multi-component mixtures is difficult. Because of this, it is strongly suggested that the virtual mass coefficient and diffusion coefficients be handled more accurately for these kinds of flows. The values of these coefficients significantly affect how accurate proposed TFM predictions are, but there hasn't been much research on how to estimate them. The study also sheds light on the model's accuracy and highlights the areas where the model's predictions will be mathematically trustworthy. • Evaluate SIMMER code for ill-posedness in a two-component, two-phase flow. • Study the effects of multiple parameters on continuity equation ill-posedness. • Explain key mathematical characteristics in liquid-metal and gas two-phase flow modeling. [ABSTRACT FROM AUTHOR]