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With nuclear energy’s rapid development in recent years, supply of nuclear fuel has become increasingly important. Thorium has re-gained attention because of its abundant reserves and excellent physical properties. Compared to the homogeneous Th/U MOX fuel, separation of thorium and uranium in space is a better use of thorium. Therefore, this paper describes four-level spatial separation – no separation, tristructural-isotropic (TRISO) level, channel level and block level – in a block-type thorium-loaded helium-cooled high-temperature reactor (HTR). A traditional two-step calculation scheme, lattice calculation followed by core calculation, is used to get the neutronic performance of the equilibrium cycle, including uranium enrichment, mass of fuel, effective multiplication factor, and average conversion ratio. Based on these data, the fuel cycle cost of different-scale spatial separation can be calculated by the levelised lifetime cost method as a function of thorium content. As the separation level increases from no separation to channel level, the effective enrichment decreases 15% due to the increase of resonance escape probability. So there is a 13% drop for the fuel cycle cost. For TRISO-level separation, as the thorium content increases from 9 to 57%, the effective enrichment decreases 14% because of the superior breeding capacity of U-233. As a result, the fuel cycle cost also has about a 12% decrease. From the perspective of fuel cycle economics, channel-level separation with 60% thorium content is suggested. [ABSTRACT FROM AUTHOR]

(Very) high temperature gas-cooled reactors have a reactor outlet temperature of 900–1000 °C, which provides a base for higher cycle efficiency by adopting closed nuclear gas-turbine combined cycles (NGTCCs) instead of direct regenerative cycles (RCs). Typical NGTCCs (TCCs) usually only adopt precoolers and do not include recuperators. The analysis of precoolers in the TCC indicates that a simplified TCC (SCC), a TCC without precoolers, has higher cycle efficiency because of a higher reactor inlet temperature. Moreover, a criterion whether a TCC adopts precoolers is derived. If NGTCCs includes recuperators, it becomes a typical compound combined cycle (TCCC). The analysis of recuperators in the TCCC indicates that recuperator’s effectiveness should be above 0.7 for high efficiency, otherwise recuperators can only be used to adjust the inlet temperatures of reactors and steam generators. Based on the discussion of precoolers and recuperators in NGTCCs, a simplified TCCC (SCCC) that only including recuperator is proposed. The further parametric calculations show that the optimum cycle efficiency is 50.0–57.8% for RC, 51.5–59.2% for SCC and 53.3–61.5% for SCCC, with a reactor outlet temperature of 900–1200 °C. Once the reactor inlet temperature is limited to 400–750 °C, a threshold value (500–600 °C) of reactor inlet temperature that separates the working condition of SCC and SCCC is obtained. The SCC is most favorable when the limit temperature is below the threshold value; otherwise, the SCCC has the highest efficiency. [ABSTRACT FROM AUTHOR]

• Detailed description on the governing equations and key physical phenomena in the fuel pellet and cladding are summarized. • Mature and next generation fuel performance analysis codes are concluded. • Multi-physic coupled analysis codes are focused for the fuel performance analysis. • Especially, the physical process and key phenomena are concluded during LOCA, RIA and severe accidents. Nuclear fuel performance analysis is fundamental for the design of pressurized water reactors (PWRs) with great safety concerns. The analysis of a fuel rod involves many complex processes, including heat transfer, mechanics, material, mass transfer and chemistry interaction. This paper reviews the basic frame and recent progress of fuel performance analysis of PWRs during normal operation and accidents. Three conservation governing equations of heat, momentum and mass transfer, and the solution strategies are mentioned first. Reviewed phenomena include gas gap heat transfer, fission gas release, fuel pellet relocation, pellet cladding mechanical and chemical interactions, waterside corrosion, etc. Tools used to execute the performance analysis, namely, fuel performance analysis codes, are divided into mature and next generation codes and typical codes are also reviewed. FRAPCON/FRAPTRAN and FROBA are introduced as two representatives of mature generation codes, and BISON, ALCYONE and CAMPUS are reviewed as the representatives of the developing generation codes. The fuel performance analysis in design-basis and severe accidents are also extensively reviewed as a new developing topic of fuel performance analysis, particularly in severe accidents. [ABSTRACT FROM AUTHOR]

• The effects of different geometrical parameters on flow and heat transfer in the extended domain are studied. • The influence of the extended domain is explained from the angle of relative pressure and pressure gradient. • The criterion for judging the accuracy of numerical simulation results of the extended domain is put forward. • The optimal geometric parameters of the extended domain are selected. The natural convection flow in the vertical rectangular channels is frequently encountered in many engineering applications. Currently, most relevant simulation studies neglected the influence of the extended domain on the calculation results, it could cause the heat transfer coefficient to deviate by more than 10% from the actual result. This paper aims to study the influence of the extended domain on the natural convection heat transfer in a vertical rectangular channel. It is found that the mass flow rate in the channel can rise by approximately 8% and the heat transfer coefficient by up to 13.5% with an increase in the extended domain. This influence is explained from the angle of relative pressure and pressure gradient. The influence of different geometric parameters on the extended domain and the selection of the optimal geometric parameters of the extended domain are discussed. Furthermore, it is expanded to different heated boundary conditions and geometric structures of the channel. This study provided a reference for the selection of extended domains in the corresponding natural convection. [ABSTRACT FROM AUTHOR]

• A coupled OpenMC/OpenFOAM code system has been developed and applied to the VHTR fuel pebble. • The coupled code system based on Monte Carlo and CFD method includes the TRISO double heterogeneity and multiscale effects. • The neutronic and thermal coupling effect of a VHTR fuel pebble was evaluated. • The traditional maximum temperature calculation method without neutronic feedback is slightly conservative. Pebble bed very high-temperature gas-cooled reactor (VHTR) has the advantages of inherent safety and high outlet temperature. It adopts a pebble fuel element with a double heterogeneity structure. This paper presents the thermal evaluations of the VHTR fuel pebble using the OpenMC/OpenFOAM code system. Firstly, the neutronic model using Cell-Based Method (CBM) and the multiscale heat conduction model are established respectively. The multiscale heat conduction model decomposes the fuel pebble on two different scales: 1) the mesoscale, defined as the fuel pebble scale; and 2) the microscale, defined the TRISO fuel particle scale. The neutronic and thermal (N/T) calculations are coupled by Picard method. The results of the steady-state coupling study of the VHTR fuel pebble show that the spatial self-shielding effect has a greater impact on the temperature of the center of the fuel pebble compared with other locations. For different fuel pebble power and pebble boundary wall temperature, the maximum mesoscale temperature and microscale temperature calculated with N/T coupling are lower than those calculated without N/T coupling. Therefore, the traditional maximum temperature calculation method without N/T coupling is conservative. And the conservative degree increases with the operating parameters of the fuel pebble. [ABSTRACT FROM AUTHOR]

• A geometrically exact coupled OpenMC and OpenFOAM model has been developed for VHTR fuel pebbles. • The effective thermal conductivity (ETC) of fuel pebbles was evaluated based on a neutronic and thermal (N/T) method. • The peak temperature calculated by traditional multi-scale heat conduction methods is conservative. • The parametric study of ETC for VHTR fuel pebbles under N/T coupling method was carried out. Pebble bed very high-temperature gas-cooled reactor (VHTR) has many economic and safety advantages, such as high safety and high outlet temperature. It uses fuel pebbles with Tristructural-isotropic (TRISO) fuel particles. The effective thermal conductivity of fuel pebbles is a key parameter to evaluate the peak temperature and average temperature of the fuel pebbles. The currently available effective thermal conductivity models do not consider the internal heat sources and the complex neutronic and thermal (N/T) coupling effects in the fuel pebbles. In this study, a three-dimensional fuel pebble heat transfer and a Monte Carlo neutronic coupled model were developed. The model has a random distribution of TRISO fuel particles and an accurate geometric model. The results show that traditional multi-scale heat conduction methods based on the existing effective thermal conductivity model, such as those based on the Chiew and Glandt models, are conservative in predicting the temperatures of VHTR fuel pebbles. The degree of conservatism is affected by fuel pebble power and packing fraction. Under the N/T coupling condition, the effective thermal conductivity is not much affected by the power of the fuel pebble but decreases with the increase of the fuel pebble boundary temperature. Geometric parameters affect the internal temperature distribution and peak temperature by changing the uniformity of the heat source distribution. [ABSTRACT FROM AUTHOR]

• Predicted the occurrence position and power of the rod bundles' CHF. • Compared the influence of cladding materials' properties on the rod bundles' CHF. • Provided a detailed analysis of the impact of grid spacers on the rod bundle' CHF. The critical heat flux (CHF) is one of the safety standards for core thermal design and is crucial for the normal operation of the reactor. The CHF problem of rod bundles with grid spacers is a hot topic in related research. The successful design of grid spacers can improve channel CHF by affecting the flow of downstream coolant, and this detection analysis can be achieved through numerical simulation methods. In addition, most simulation studies have not considered the solid region of the cladding, so there is little research on the effect of cladding materials on CHF. The role of grid spacers in rod bundle CHF is the focus of this paper, and the influence of cladding materials is another discussion direction. In this paper, following meticulous mesh generation and computational model validation, a study on the CHF of the 5 × 5 rod bundle with multiple grid spacers and different cladding materials is carried out by using CFD software. The verification results indicated that the physical model established in this study can reasonably predict the generation location and critical power value in the 5 × 5 rod bundle with grid spacers. Furthermore, the study identified that the grid spacer exerts a significant influence on coolant flow within the channel, while the weakening of the mixing effect of the grid spacer on the coolant flow is the main factor to lead the boiling crisis upstream of the grid spacer. By comparing the effects of thermal properties of different materials on CHF, it was found that when the thermal conductivity ratio of the two materials is 1.08, CHF is almost identical under the condition of a reference heat flux of 0.1 MW/m2 as a step size. In addition to thermal properties, further attention may be paid to the influence of surface characteristics of materials on CHF in future work. [ABSTRACT FROM AUTHOR]

• The transient thermal hydraulic performance of the Secondary Side Passive Residual Heat Removal Heat Exchanger (SPRS HX) in the HPR1000 reactor is simulated using computational fluid dynamics (CFD). • The C-shape tube bundle is modeled using the porous media approach along with the distributed resistance method. • The two-phase flow phenomenon in the cooling water tank (CWT) is simulated by the drift flux model. • Three dimensional distributions of fluid temperature, velocity and void fraction in the CWT are obtained. • The SPRS HX heat transfer capacity from the primary side to the secondary fluid is determined. In advanced pressurized water reactor HPR1000, the secondary side passive residual heat removal heat exchanger (SPRS HX) is a crucial component. It removes the decay heat using boiling and convection heat transfer in the event of accidents. The robust functioning of the SPRS HX significantly influences overall reactor safety. To investigate the thermal hydraulic attributes of SPRS HX, computational fluid dynamics (CFD) is used in this study. The tube region is simulated using the porous media method (PMM). To model the two-phase flow in the tank, the drift flux model (DFM) is employed. The assessment of heat transfer rate from the primary to the secondary sides is conducted by empirical correlations. The governing equations are addressed through the utilization of FLUENT. Specifically, the momentum and energy equations are augmented with additional source terms that account for flow resistance and heat transfer in the tube region. These modifications are enacted through User Defined Functions (UDF). Three-dimensional profiles of the void fraction, fluid temperature and velocity are acquired, allowing for a comprehensive analysis of the heat transfer attributes of tube region. [ABSTRACT FROM AUTHOR]

• A new method of fuel pebble temperature calculation based on neutronic and thermal coupling equivalent model is introduced. • The new method is evaluated on steady-state and transient multi-scale neutronic and thermal hydraulic (N/TH) coupling calculation. • The traditional temperature calculation method, not considering N/TH coupling at fuel pebble scale, is conservative. As one of the six reactor types in the fourth-generation nuclear energy systems, the Very High Temperature Gas-cooled Reactor (VHTR) possesses inherent safety features. One VHTR fuel pebble contains tens of thousands of randomly distributed TRISO fuel particles, displaying a double heterogeneity characteristic. Current research models employ the homogenized method for fuel pebble temperature calculation, which neglected the neutronic and thermal (N/T) coupling effect and double heterogeneity characteristic. This study proposes a fuel pebble coupled equivalent method (PCEM) for temperature calculation. The method is based on the effective thermal conductivity obtained by the fine structure fuel pebble N/T coupling method. The new method is evaluated on steady-state and transient VHTR multi-scale neutronic and thermal hydraulic (N/TH) coupling calculation. The results show that the traditional temperature calculation method is conservative. Under steady-state conditions, the maximum degree of conservatism of the fuel pebble average temperature and peak temperature are 0.6 K and 14.2 K respectively. For the sudden drop in coolant flow rate transient condition, the average temperature and peak temperature of the fuel pebble calculated by PCEM are smaller than the values calculated by traditional method at the same time. Due to the influence of the fuel Doppler effect, the transient response of the traditional calculation method is slower. [ABSTRACT FROM AUTHOR]

• The strongly coupling relationship between curvature and inclination is discovered. • The inflection point of inclination effect on condensation characteristics is found. • The mechanism of coupling effect between inclination and curvature is revealed. • The condensation performance of various curvature and inclination is compared. Steam-air condensation is one of the critical thermal–hydraulic phenomena in containment under accident conditions. Currently, sufficient experimental research and numerical analysis have been conducted on steam–air condensation outside the tube/plate. However, few studies have investigated the influence of curvature and inclination on condensation characteristics, and the local phenomena and mechanisms that affect the condensation process have not yet been fully revealed. In addition, it is of vital importance to evaluate the coupling effect between curvature and inclination in practical applications. To investigate the above-mentioned effects, numerical simulations were conducted over a wide parameter range (P = 0.2 ∼ 1.6 MPa, ω a = 0.1 ∼ 0.9, ΔT = 20 ∼ 100 K), and the related mechanisms were analyzed in detail. The results showed that there is a strong coupling relationship between curvature effect and inclination effect and the inflection point in the influence of inclination effect on steam condensation characteristics is discovered. Compared to the traditional conclusion that condensation capability gradually increases with the decrease of inclination angle. The effect of inclination on the condensation characteristics can reverse in different curvature regions. When the curvature is greater than 5 m−1, the inclination angle is negatively correlated with the condensation capability. When the curvature is less than 5 m−1, the inclination angle is positively correlated with the condensation capability. Furthermore, the enhancement effect originate from the curvature effect will greatly decrease as the inclination angle increases. [ABSTRACT FROM AUTHOR]

• A Monte Carlo based generation method of multigroup cross sections for pebble type VHTR by OpenMC was developed. • The new method can effectively deal with the double heterogeneity effect of pebble type VHTR. • Two different level models are proposed to be used to assess the accuracy of the cross section generation method. • Neutronics calculation results using multigroup cross sections by homogenized model was obtained. Very high temperature reactor (VHTR) is a Generation-IV (Gen-IV) type of reactor which could nominally operate at a high temperature. In traditional two-step core neutronics calculation method of VHTR, the high precision multigroup cross sections are required. Therefore, a new Monte Carlo based generation scheme for multigroup cross sections of pebble type VHTR, based on open source code OpenMC, is proposed in this paper. The new method can effectively deal with the double heterogeneity effect of pebble type VHTR. And the generated multigroup cross sections can be applied to the deterministic neutronics calculations and OpenMC multigroup calculations. Moreover, a unit cell model and a simplified pebble bed lattice model of VHTR were developed to be used for verification of the method. According to the verification results, the new developed method using OpenMC to generate multigroup homogenized cross sections for pebble type VHTR is feasible and enough accurate. [ABSTRACT FROM AUTHOR]

• Compared with tight coupling, loose coupling dominates in the N/TH coupling studies of nuclear reactors. • The N/TH loose coupling methods are mainly categorized into four types based on 185 papers. • The performance improvement and application studies of the OS, Picard iteration and JFNK methods are concluded. • Typical neutronic and thermal-hydraulic codes and coupling codes are presented. In an operating nuclear reactor system, various physical phenomena of different properties are intimately linked. These multiphysics phenomena include neutronics (N), thermal-hydraulics (TH), materials science, and other subjects. Among them, the interaction between neutronics and thermal-hydraulics is of great significance in reactor design and safety analysis. In this work, different N/TH coupling methods are reviewed, including loose and tight coupling. For the studies on loose coupling, in which two physical fields are decoupled, the current research status is summarized and classified based on the coupling methods of neutronics and thermal-hydraulics. The studies of tight coupling are introduced based on multiphysics coupling platforms. The investigation shows that the number and objectives of loose coupling studies are more abundant and extensive than those of tight coupling. This indicates that loose coupling strategies are the mainstream coupling solutions in recent research. Furthermore, the solution approaches of N/TH coupling are reviewed with respect to the aspects of performance improvement and application studies, including the operator splitting (OS), Picard iteration, and Jacobian-Free Newton–Krylov (JFNK) methods. A comprehensive study of the solution approaches shows that most of the current loose coupling numerical simulations adopt the Picard iteration method, because it has higher calculation accuracy than the OS method. In contrast to the decoupling approaches such as the OS and Picard iteration methods, the JFNK method updates all physical quantities synchronously, which makes it more accurate. Hence, there are broad application prospects for N/TH tight coupling of the JFNK method. [ABSTRACT FROM AUTHOR]

Highlights • The performance of the thorium MOX fuel blocks with four type driver fuels in AHTR is compared. • The Th/U233 achieves the highest k inf in the BOL. • The fuel blocks with Pu-239 as a fissile isotope have less k inf and higher CR than Th/WGrU. • Th/RGrPu achieves the maximum discharged burnup which allows for a longer operation period and completes plutonium fissions. Abstract The use of thorium-based fuels is an attractive option to achieve high burnup level and relatively longer reactor operation period. This also includes some other potential advantages, such as, low proliferation risk and high conversion ratio, etc. It is interesting to utilize thorium-based fuels in advanced high temperature reactors (AHTRs), because they are a better choice for their high burnup limit and inherently-safe HTRs with a low-pressure operation condition. The neutronic performance of thorium MOX fuel blocks, including infinite multiplication factor, conversion ratio, discharge burnup and the neutron spectra in AHTRs, has been investigated by the DRAGON V4 code for four different driver fuels, U-233, reactor-grade plutonium (RGrPu), weapon-grade plutonium (WGrPu) and weapon-grade uranium (WGrU). Based on a typical fuel block of AHTRs, the influence of varying heavy metal loading and effective enrichment on the neutronic performance of the four fuel blocks is firstly investigated. The heavy metal loading of 5 kg and effective enrichment of 12% are selected for the neutronic comparisons. The further comparisons of the four thorium fuels with different driver fuels show that the Th/RGrPu and Th/WGrPu fuel blocks have intermediate neutron spectra, while Th/U233 and Th/WGrU fuel blocks have thermal neutron spectra. The Th/U233 fuel achieves the highest infinite multiplication factor at beginning of life because of its high thermal η value. Th/RGrPu fuel achieves the highest conversion ratio and the highest discharge burnup. [ABSTRACT FROM AUTHOR]

Steam condensation in the presence of air is a vital heat transfer process in postulated loss of coolant accidents. To numerically evaluate local and global thermal hydraulic properties, several single-phase models have been developed in previous studies. However, potential mutual interactions between the liquid and gas phase (e.g. evaporation, radioactive material retention, etc.) impose requirements on two-phase models. Accordingly, the present work performed CFD evaluations on a two-phase boundary layer model (TPBL) which concerned the gas phase, liquid phase and their mutual interactions. Assessments were performed by comparing calculated results with the COPAIN and TOSQAN experimental data. The COPAIN cases show that predicted local and average heat fluxes are generally within 25% and 15% deviation, respectively. Detailed properties were obtained on the formation, distribution, and thermal resistance of liquid film as well as the model computational costs. In the TOSQAN cases, two steady states, including steam-air mixtures and steam-air-helium mixtures were considered. Via these cases, the applicability of the TPBL model in predicting local field profiles like velocity, temperature and concentration distribution was discussed. Results demonstrate that the calculated results overall agree well with the experimental ones. In general, the TPBL model is feasible in postulated accident analysis. [ABSTRACT FROM AUTHOR]

• Modeling and analysis of the entire small pressurized water reactor nuclear power plant were conducted using the MELCOR code. • This paper analyzed the total release quantity and distribution patterns of radioactive aerosols and inert gases. • This paper analyzed the impact of the activation of the containment spray system on the release and distribution of radioactive aerosols and inert gases. The small-sized pressurized water reactor (PWR) characterized by its high inherent safety and wide applicability, has become a promising type of reactor with broad prospects for development. Currently, there is a limited amount of research dedicated to the analysis of accident source terms specific to small-sized pressurized water reactors, this study adopts the ACP100 as a reference to establish a model of a small-sized pressurized water reactor power plant using the severe accident analysis code MELCOR. This study investigated the release and distribution characteristics of radioactive materials under the scenario of a large break loss of coolant accident in the heat pipe section. Specifically, it analyzed the impact of opening and shutting down the containment spray system on the release and distribution of radioactive materials. The results indicate that in the absence of the containment spray system, the majority of radioactive aerosols are primarily distributed in liquid form within the containment, with a small portion depositing on other components and dispersing in the gas phase; inert gases are primarily distributed in the gaseous phase within the containment. With the containment spray system engaged, there is an increase in the total quantity of radioactive aerosols and the concentration of radioactive aerosols in the gaseous phase significantly decreases, while the concentration of liquid-phase radioactive aerosols dramatically increases. The containment spray system exhibits strong capabilities for the removal and retention of radioactive aerosols. While the total release quantity remains largely unchanged, the time to reach the peak release of inert gases is significantly shortened The containment spray system significantly affects the release rate of inert gases. [ABSTRACT FROM AUTHOR]

• A neutron diffusion kinetics solver DiFOAM V1.0 is developed for pebble type VHTR based on OpenFOAM. • The finite volume method is used to solve neutron diffusion equations and delayed neutron precursor equations. • Two different scale problems including PWR assembly and core benchmark are used to be assessed the accuracy of DiFOAM V1.0. • The accuracy of OpenMC/DiFOAM V1.0 code system applied to pebble type VHTR has been proved. As an advanced reactor type, the pebble type very high temperature gas cooled reactor (VHTR) has the advantages of high safety and high outlet temperature. In this study, the OpenMC/OpenFOAM code system is developed for a neutronic analysis of VHTR. The workflow of OpenMC/OpenFOAM code system consists of two steps: the OpenMC code is used to generate few-group cross sections; then, the OpenFOAM code is used to perform core neutron diffusion kinetics analyses and a neutron diffusion kinetics simulation solver (DiFAOM V1.0) is developed. The finite volume method (FVM) is used to discretize the neutron diffusion kinetics equations. Two scale problems are selected to verified the DiFAOM V1.0 solver. Firstly, the PWR assembly model are compared by DiFAOM V1.0 and Monte Carlo results. The results show that for PWR assembly, the difference of k eff is 41 pcm, and the maximum error of four group normalized flux is −0.0841%. Secondly, 2D-TWIGL benchmark is used to assess the accuracy of DiFAOM V1.0 at core scale, and the calculated results of k eff and normalized power distribution calculated were in good agreement with the benchmark. The above results prove the correctness of DiFAOM V1.0 solver. And for application of VHTR BCC lattice model, the maximum error of four group normalized flux is only −0.826%. [ABSTRACT FROM AUTHOR]

• The pressure range for the experiment was 0.2~1.6 MPa and the subcooling was 28~114 °C. • The effect of tube diameter and inclination angle in the tube bundle was analyzed. • A segmented experimental correlation with high relevance was obtained. Steam condensation accompanying with air is an important thermal-hydraulic phenomenon in reactor containment. Condensation outside tubes is an important method to remove heat out of the containment under an accident. Most of the previous studies have been conducted on single tube and vertical plates, where the findings may not reflect the heat transfer in tube bundle well. The experimental studies were firstly carried out based on 3 × 3 tube bundle with diameters of 12–19 mm and inclination angles of 0–90° in the parameter range of pressure 0.2–1.6 MPa, air mass fraction of 12.1–75.2% and wall subcooling of 28.3–114.1℃. The results indicate that the effect of diameter and inclination angle in the tube bundle varies significantly in different pressure ranges. When the pressure is less than 0.8 MPa, the condensation heat transfer coefficient (CHTC) generally increases with decreasing tube diameter and inclination angle. The CHTC of the heat transfer tube with a tube diameter of 12 mm and an inclination angle of 0° increased by a maximum of 29% compared to that of the heat transfer tube with a tube diameter of 19 mm and an inclination angle of 90°. When the pressure is 0.8–1.6 MPa, the overall CHTC decreases as the diameter of the tube decreases, and the maximum reduction is about 20%. It decreases first and then increases with the decrease of the inclination angle, and the CHTC is the smallest when the inclination angle is about 60°. [ABSTRACT FROM AUTHOR]

• A new BP-ANN calculation method for reactor fuel reloading optimization was developed. • The BP-ANNs developed in this paper have higher precision than those in similar paper. • Five different meta-heuristic algorithms were compared when combining with the BP-ANN calculation method for fuel reloading optimization. • The developed BP-ANN calculation method for reactor fuel reloading optimization obviously fasten the calculation time. In order to obtain optimal core configurations, the problems of fuel reloading optimization must be solved in nuclear engineering field. This problem is a combinational optimization problem with an enormous solution space. In the traditional calculation method, the meta-heuristic algorithms (MHAs) are used to search and update optimization solutions continuously, while the core physics calculation programs are used to calculate the optimization parameters. In this research, we developed a new BP-ANN calculation method to solve the problems of fuel reloading optimization for thorium-based block-type high temperature gas-cooled reactors (HTGRs). In this method, the function of core physics calculation programs are replaced by BP-ANNs in order to reduce calculation time. Besides, five different MHAs, which are genetic algorithm (GA), particle swarm optimization (PSO), Teaching-learning based optimization algorithm (TLBO), harmony search (HS) and Rank-based ant system (RAS), are used under this new method to solve the problems of fuel reloading optimization, respectively. The results shows that PSO is the best MHA with a real feasible solution ratio of 0.30 and the best fitness function value of 0.455904. Compared to the traditional calculation method, the calculation time of BP-ANN calculation method is reduced to 0.7% of the former. [ABSTRACT FROM AUTHOR]

• The four flow modes were observed and analyzed of an EHDNCS. • The stability map of EHDNCS has been obtained based on 134 experimental cases. • The regions of stability and CIWH can be well divided by different equilibrium qualities. • Discrete and amplitude characteristics of the flow rate were analyzed for the four flow modes. • Increasing the inlet temperature and resistance is more easily induce flow instability of EHDNCS. As nuclear power plants tend to be miniaturized and integrated, equal-height-difference natural circulation systems (EHDNCSs) have potential applications in floating nuclear power plants. This work experimentally investigated the flow modes of an EHDNCS at low pressure. It has observed four quintessential flow modes in the experiments. The transient characteristics of each flow mode were analyzed by the visualization results. In addition, the dispersion and amplitude characteristics of flow rate were compared in different flow modes based on 134 experimental cases. The experimental results reveal that the flow oscillation is related to unsteady flow pattern for intermittent and condensation oscillations. The condensation induced water hammer (CIWH) is found in the condensation oscillation. According to the stability map, it is found that the regions of stability and CIWH can be well divided by equilibrium quality of −0.23% and 0.47%, and the unstable range is enlarged by increasing the inlet temperature and resistance. [ABSTRACT FROM AUTHOR]

• The performance of the thorium seed and blanket (S&B) fuel blocks with four type driver fuels in AHTR is investigated. • The discharge burnup is the highest for the fuel block with U-233 as driver fuel. • The fuel block with RGrPu as driver fuel has the highest conversion ratio. • The temperature coefficients of reactivity are more negative for RGrPu and WGrPu driven fuel blocks. Because of the high burnup achieved by thorium fuels, they are attractive for utilization in the so-called advanced high temperature reactor (AHTR), which is a new reactor that merges HTR and molten salt reactor (MSR) designs. Based on a typical fuel block of AHTR, the neutronic characteristics of seed and blanket (S&B) fuel blocks with different driver fuels, including the infinite multiplication factor (k inf), conversion ratio (CR), discharge burnup, energy spectra, equilibrium mass of U-233 and temperature coefficient of reactivity (α T), were investigated based on DRAGON V4. The seed is assumed to contain thorium with either U-233, RGrPu, WGrPu or WGrU while the blanket is always thorium. In terms of maximizing discharge burnup, 6 kg heavy metal (HM) per fuel block was selected for the comparison analysis of the four drivers. The same configuration of S&B 4 + 5 was used for all the investigated fuel blocks. It is shown that the k inf has the highest value for the fuel block with U-233 as driver fuel (Th/U233) and decreases passing to Th/WGrU, Th/WGrPu and Th/RGrPu. Though Th/RGrPu achieves the highest CR, Th/U233 achieves the deepest discharge burnup because of its high k inf. The temperature coefficients of reactivity of fuel (α TF) and moderator (α TM) of Th/RGrPu, Th/WGrPu, and Th/WGrU are all negative. However, α TF of Th/U233 fuel block is negative but its α TM is positive. [ABSTRACT FROM AUTHOR]