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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]

Summary: Numerical simulation of the Space Reactor Power System is of great importance to reactors development and application in aerospace. In this paper, thermal‐hydraulic and thermoelectric characteristics analysis based on 1/6th full thermionic reactor core of the TOPAZ‐II were performed using 3‐D computational fluid dynamics software STAR‐CCM+. A Benchmark study was conducted by comparing the numerical results with experimental data and design values with a maximum error of 5.4%‐11.8%. For the thermal‐hydraulic analysis: (a) the highest temperature point occurred at the center hole of the center thermionic fuel element (TFE) with the temperature of 2303 K; (b) the maximum temperature difference located in the fuel region and interelectrode gap with the value about 200 K and 1000 K, respectively; (c) axial temperature distribution of coolant is similar in each flow channel, and the maximum temperature difference appears at the reactor core outlet with the value of 10 K. For the thermoelectric analysis, the potential on center or the edge TFE is much higher or lower than the others, the calculated electrical power of the full thermionic reactor core is 4.95 kW, the distribution trend of current and potential is in a good agreement with the axial emitter temperature. [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]

Thermal-hydraulic performance in the reactor core was an essential factor in the nuclear power plant design. In this study, we analyzed the three-dimensional (3-D) thermal-hydraulic characteristic of reactor core based on porous media method. Firstly, a 3-D rector pressure vessel (RPV) model was built, including the inlet leg nozzle, downcomer, lower plenum, reactor core, upper plenum and outlet leg nozzle. Porous media model was used to simplify the reactor core and upper plenum. The commercial CFD code ANSYS CFX was employed to solve the governing equations and provide the 3-D local velocity, temperature and pressure field. After appropriate parameters and turbulent model being carefully selected, the simulation was validated against the 1:5 scaled steady-state hydraulic test. The predicted hydraulic parameters (normalized flowrate distribution and pressure drop) were in good agreement with the test results. And the predicted thermal parameters agreed well with the designed values. The validation indicated that this method was practicable in analyzing the 3-D thermal-hydraulic phenomena in the RPV. Finally, the thermal-hydraulic features in reactor core were analyzed under the condition of the Steam Generator Tube Rupture (SGTR) accident. The simulation results showed that the coolant temperature increased gradually from the center to the periphery in the reactor core in the accident. But the temperature decreased to safety level rapidly after the reactor shutdown and safety injection operation. The reactor core could keep in a safe state if appropriate safety operations were performed after accidents. [ABSTRACT FROM AUTHOR]