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• Concept of "LESS is MORE" is proposed in the design of lead-bismuth cooled reactor. • Core design of lead-bismuth cooled small modular reactor is conducted. • Neutronic and thermal-hydraulic performances of reactor core are analyzed and improved. • Preliminary core design of LBE cooled reactor could satisfy the design criteria and is reasonable. Small modular reactors (SMR) are in great demands in the near future and the lead-based fast reactor is one of the most important directions in the development of SMR. In this paper, core design and analysis of a LEad-bismuth Small MOdular Reactor (LESMOR) was developed and examined to realize the notion of "LESS is MORE". LESMOR is a 300 MWt pool-type reactor where all facilities are mounted in a reactor vessel. Advanced nitride fuel is adopted, and the spent-plutonium is used as the driven fuel and thorium is used as the fertile fuel. Sub-channel analysis was performed in the assembly design using an in-house sub-channel code SUBAS, and the 11 × 11 scheme with a pitch-to-diameter (P/D) ratio of 1.4 was adopted. The full core neutronic performance was studied and evaluated in this paper. Results show that the reactor can be maintained critical for 20 years and the maximum radial power factor is 1.28. The steady-state thermal-hydraulic performance of the core was analyzed, including the characteristics of the mass flow distribution and the hottest assembly. The peaking fuel cladding temperature is 519.2 °C, and the maximum fuel center temperature is 723.4 °C, both within the temperature limit. This paper provides valuable information in the neutronic and thermal-hydraulic design and analysis of LBE nuclear reactor. [ABSTRACT FROM AUTHOR]

• An innovative three-dimensional thermal–hydraulic code, CorTAF, was developed based on OpenFOAM. • The CorTAF is more computationally efficient with the spatial resolution of subchannel level. • The CorTAF is validated against experiment benchmarks and subchannel analysis codes. • The steady state and transient simulations of a full scale PWR core were carried out using CorTAF. In this paper, a thermal hydraulic analysis method based on open source CFD platform OpenFOAM was proposed. The models of coolant flow and heat transfer, fuel rod heat conduction and coupled heat transfer were established according to the bundle structure characteristics of PWR core, and then a nuclear reactor three-dimensional thermal–hydraulic characteristics analysis code CorTAF was developed based on finite volume method with the spatial resolution of subchannel level. The international benchmarks, including GE3 × 3, Weiss and PNL2 × 6 fuel assembly flow and heat transfer experiments, were selected to carry out the validation. The calculation results were basically consistent with the experimental data, illustrating that CorTAF is suitable for predicting the flow and heat transfer characteristics of coolant in the rod bundle fuel assembly. The thermal–hydraulic characteristics of fuel assembly and PWR core under full power operation and blockage accident conditions were simulated using CorTAF. The steady-state and transient spatial distribution of physical quantities in subchannel-scale including coolant and fuel rod temperature were obtained, and the influence of blockage condition on flow and heat transfer characteristics was analyzed. This work has reference significance for the development of core thermal hydraulic analysis tools for PWR. [ABSTRACT FROM AUTHOR]

Fuel pins assembled in Pressurized Water Reactor (PWR) and Liquid Metal-cooled Fast Reactor (LMFR) core are usually cooled by single-phase coolant such as water or liquid metal. Researches on the coolant flow and heat transfer phenomena through rod bundles is quite essential for the nuclear reactor core design. This paper briefly summaries the current developments of Computational Fluid Dynamics (CFD) studies on the thermal hydraulic analysis of single-phase flow in the rod bundles of nuclear reactor core. On account of the tremendous amount of publications related to each aspect above, this brief review would focus on the most representative researches in the past decades. The aim of this paper is to provide an outline and present a comprehensive overview of the CFD research and guidelines on single-phase flow through bare rod bundle rod with mixing vane spacer and wrapped wire spacer in reactor core. In addition, the analysis and validation methods for various CFD calculating methods are concluded. The applications of CFD approach with different type of rod bundles are summarized. Finally, some remaining challenges for future work are proposed and discussed. [ABSTRACT FROM AUTHOR]

Summary: The development of deep space exploration requires space nuclear reactor with higher power and better endurance. In order to meet the increasing high power requirements for space applications, an Open‐grid MEgawatt Gas‐cooled spAce nuclear reactor (OMEGA) is proposed in the paper, which is featured with low specific mass and high energy conversion coefficients. A set of models, including neutron kinetics model, reactivity feedback model and heat transfer model are established. Besides, the space nuclear system analysis code (System Analysis code of MEgawatt gas‐cooled space Reactor [SAMER]) is developed by finite volume method and staggered grid technique for the preliminary safety analysis of the OMEGA conceptual design. Furthermore, the calculation results provided by the SAMER code agree well with the computational fluid dynamics simulation, validating the reliability of the code. Moreover, the transient thermal‐hydraulic responses of the OMEGA under the unprotected reactivity insertion accident (URIA) and the unprotected loss of partial coolant flow accident (LOFA) are obtained and analyzed. The maximum fuel temperature after the URIA has a margin of 262 K from the fuel temperature limit value. The total power after the LOFA is about 68.7% of the rated power, meanwhile, the fuel radial temperature distribution tends to be more uniform with the fuel average temperature unchanged. It can be found that the OMEGA design shows satisfactory safety characteristic and reliability under the URIA and the LOFA scenarios. This paper is the stage summary of the conceptual design work of the OMEGA, and it may provide useful reference for megawatt space nuclear system design and thermal‐hydraulic analysis. [ABSTRACT FROM AUTHOR]

The thermal hydraulic and neutronics coupling analysis is an important part of the high-fidelity simulation for nuclear reactor core. In this paper, a thermal hydraulic and neutronics coupling method was proposed for the plate type fuel reactor core based on the Fluent and Monte Carlo code. The coupling interface module was developed using the User Defined Function (UDF) in Fluent. The three-dimensional thermal hydraulic model and reactor core physics model were established using Fluent and Monte Carlo code for a typical plate type fuel assembly, respectively. Then, the thermal hydraulic and neutronics coupling analysis was performed using the developed coupling code. The simulation results with coupling and noncoupling analysis methods were compared to demonstrate the feasibility of coupling code, and it shows that the accuracy of the proposed coupling method is higher than that of the traditional method. Finally, the fuel assembly blockage accident was studied based on the coupling code. Under the inlet 30% blocked conditions, the maximum coolant temperature would increase around 20°C, while the maximum fuel temperature rises about 30°C. The developed coupling method provides an effective way for the plate type fuel reactor core high-fidelity analysis. [ABSTRACT FROM AUTHOR]

Summary: A thermoelectric analysis of space nuclear power reactor is of great importance to the space reactors development. In this paper, the reactor core model of TOPAZ‐II designed by Soviet Union, including neutronics model, thermal‐hydraulic model, and electrical circuit model, is established based on reasonable assumptions. A system analysis code is developed to analyze the thermoelectric characteristics of the TOPAZ‐II under the condition of steady‐state operation, start‐up procedure, power change mode, and reactor shutdown. The code has been benchmarked with experimental data, and the maximum relative error is 16.6%. Numerical results show that for the steady state, the simulated electrical power is 5.2 kW within the design value range. For transient state, the reactor thermoelectric characteristics are mainly affected by the electrode temperature: (a) For the start‐up procedure, when the emitter temperature increases above 1700 K, electrical system begins to work and reach full power in 5 minutes. (b) For power change mode, the emitter temperature decreases by 5%, while the electrical power decreases by 67%; (c) For reactor shutdown, electrical power reduces to 0 kW as the emitter temperature decreases to 1400 K in 100 seconds. This study provides valuable theoretical supports for the design and analysis of the thermionic space nuclear power reactor. [ABSTRACT FROM AUTHOR]

• A new boron diffusion model is proposed for the nuclear reactor core analysis. • The effect of turbulent diffusion is taken into consideration. • The model is implemented in the subchannel analysis code. • The CFD method is used to verify the new model. The diffusion characteristic of borated water in nuclear reactor core is important for the reactivity control during some possible reactivity transient scenarios. The non-uniform distributions of boron concentration increase the risk of reactor accidents and lead to some reactor safety problems. Hence it is necessary to develop an effective boron diffusion model into the reactor thermal-hydraulic analysis code. In this paper, a quasi-three dimensional boron tracking model, in which the turbulent diffusion effect was considered carefully, was proposed and successfully implemented in the subchannel analysis code. The reliable CFD method with k-omega turbulent model validated against International Standard Problem (ISP No. 43) experimental data was used to verify the feasibility of proposed boron tracking model. A typical Pressurized Water Reactor (PWR) 3 × 3 fuel bundle model was established and the borated water was injected into one specified channel and pure water into other channels. The distributions and variations of boron concentration along the flow direction were achieved and analyzed. Results indicate that the proposed boron tracking model could predict the boron diffusion process in the fuel rod bundles more accurately compared with the original model. This work provides an essential supplement for the nuclear reactor subchannel analysis method. [ABSTRACT FROM AUTHOR]

• The thermal behavior of a horizontal high-temperature heat pipe under motion conditions are experimentally investigated. • The motion conditions include transitory inclination, vertical heave, and periodic swing. • The impacts of motion types on temperature variation and heat transfer deterioration of heat pipes are summarized. High temperature heat pipe is widely applied to multiple industrial fields, including in small modular reactors to cool the reactor core, or as heat exchangers. The thermal behavior of the high temperature heat pipe has been widely researched by theoretical and experimental approaches, including designing, manufacture, numerical analysis, and experimental test on its thermal characteristics of start-up process and operation. As the heat pipe has a potential application protest on mobile nuclear reactors, this paper carried out experimental study on the thermal behavior of the high temperature heat pipe under various motion conditions. The temperature distributions on the heat pipe surface under transitory inclination, vertical heave, and periodic swing, are measured to evaluate the impact of motion conditions on the heat transfer of heat pipe at horizontal position. The transitory inclination causes an obvious temperature variation at both evaporator and condenser. The vertical heave, of which the motion direction is perpendicular to the heat pipe, has negligible impact on the isothermality of the heat pipe. The periodic swing leads to periodic temperature fluctuation, of which the amplitude increase with the decease of the period time. Besides, the dry out phenomenon is easier to happen at the evaporator when the condenser is longer. This paper summarizes the impact of different motion types on the thermal behavior of a horizontal heat pipe and provides a reference for the design and application of the high temperature heat pipe used in nuclear reactors. [ABSTRACT FROM AUTHOR]

• A code is developed for studying the operating thermal-hydraulic characteristics of the reactor core, which consists of core thermal-hydraulic model, neutron kinetics model and heat pipe startup model. • In order to verify the code, a series of experiments about potassium heat pipe are operated. • Steady-state and startup transient are performed by the code. • This research provides valuable experience for designing and formulating control strategies of the heat pipe cooled reactor. Heat pipe cooled reactor has been proposed for applying in the space station and underwater exploration featured with high reliability, low noise level and compact structure. A typical design of a heat pipe cooled reactor system is HP-STMCs designed by the University of New Mexico. It is conceptually designed as a fast reactor, lithium-filled heat pipes that transfer heat to potassium heat pipes radiator and thermoelectric generators. In this paper, a code is developed for studying the operating thermal–hydraulic characteristics of the reactor core, which consists of core thermal–hydraulic model, neutron kinetics model and heat pipe startup model. In order to verify the code, a series of experiments about potassium heat pipe are operated. Steady-state and startup transient are performed by the code. The highest temperatures of fuel and matrix in reactor core show at the 2nd channel, which reach 1720 K and 1545 K from the results of steady-state. As for the startup transient, the code for heat pipe cooled reactor core successfully predicts key parameters during startup transient, such as each layer material temperature of reactor core, a single heat pipe's temperature response and heat rejection power. The code has initially predicted the transient response of the heat pipe cooled reactor under power increasing rates of 800 W/s and130W/s. It is discussed that the startup strategy of power increasing rates is based on heat pipe transition temperature (916 K). This research provides valuable experience for designing and formulating control strategies of the heat pipe cooled reactor. [ABSTRACT FROM AUTHOR]

The lead-based reactor is one of the fourth-generation nuclear energy systems with excellent development prospects. It has attracted much attention because of its unique advantages in fuel transmutation and breeding, passive safety, and economic efficiency. In this study, the main features of typical lead-based fast reactors (LFRs) worldwide are reviewed and current challenges in their development are pointed out. Moreover, the research progress in four main thermal-hydraulic aspects is overviewed and discussed, including the flow and heat transfer characteristics of the coolant, thermal-hydraulic analysis of the reactor core, lead pool and reactor system. These aspects are further divided into specific issues according to the developing needs of LFR. For each issue, up-to-date research findings, existing difficulties, and future trends are presented. This study can serve as a reference for the design and thermal-hydraulic analysis of LFRs. • Research on the main thermal-hydraulic features of LFRs is discussed. • Flow and heat transfer characteristics of lead-based coolants are summarized and compared. • Up-to-date research and developments of Core/Pool/System Level thermal-hydraulics are reviewed. [ABSTRACT FROM AUTHOR]