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Summary: To meet the requirements of the unmanned underwater vehicles (UUVs) for the power source, a conceptual design of a 100‐kWe level Nuclear Silent Thermal‐Electrical Reactor (NUSTER‐100) is proposed. The NUSTER‐100 is a 1‐MWt reactor that has been designed to couple with a 100‐kWe thermoelectric generator system. The high temperature sodium heat pipes are adopted to cool the reactor core. The cascaded wide temperature range thermoelectric generation (TEG) system is used to achieve thermoelectric conversion, which ensures high power generation efficiency and low operating noise. The neutronic and thermal‐mechanical analysis results are provided for the design. The core power distribution and energy spectrum are obtained. The neutronic analysis results are imported into ANSYS as input for thermal‐mechanical analysis. In this paper, the fuel‐HPs channel and the 1/8 symmetrical core are selected as the analysis objects, and the temperature distribution and statics parameters under the normal operation condition are obtained. The analysis results show that under normal operation conditions, the core parameters meet the design requirements. These design parameters are all lower than the safety limit. This work provides a certain reference for the design and application of the heat pipe reactor power system. [ABSTRACT FROM AUTHOR]

Summary: Closed type assembly leads to complicated core thermohydraulics which are significant for design and safety analysis in sodium‐cooled fast reactor (SFR). Most of the major phenomena can be dealt with by general parallel channel network and two‐dimensional model for inter‐wrapper flow (IWF) in system code transient thermal‐hydraulic code for analysis of SFR (THACS). A main deficiency left is the overestimation of inter‐assembly heat transfer effect due to lump temperature message that one‐dimensional assembly model can only provide. A methodology was thus developed to compute wall‐subchannel temperature for simulating energy transportation through duct wall. Intra‐assembly thermohydraulics were classified as two mechanisms and non‐dimensional parameter was introduced for characterizing each of them. Ratio between linear heat flux density through hexagonal wrapper surface and pin surface qw/p was employed to account for radial heat transfer. Flow redistribution effect arouse in mixed and natural convection was concretized by a importance symbol of buoyancy force, Richardson number Ri. Two criteria were also proposed for dividing flow state into three parts and specific model for each section was defined. Two validation cases against 3‐D computational fluid dynamics (CFD) simulation of China experimental fast reactor whole core and test data from plant dynamics test loop with direct heat exchanger experiment were performed with THACS. Noticeable improvement after incorporation of this model demonstrated its value in system analysis. Although the procedure for parameter computations is geometry related, this method can be reproduced to any type of assembly directly. [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]

• 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]

• Validation of THACS and OpenMC for SFR has been enhanced by simulating FFTF LOFWST Test #13. • A new modeling method multi-parallel channel plus local subchannel model is proposed for reactor core. • Important local phenomenon like inter wrapper flow, thermal stratification is analyzed by the specific model. In order to support collaborative efforts within international partnerships on the development of sodium-cooled fast reactor (SFR), the International Atomic Energy Agency (IAEA) proposes a coordinated research project (CRP) about the unprotected loss-of-flow-without-scram test (LOSWOS) performed at Fast Flux Test Facility (FFTF) in 2018, which could validate simulation tool and numerical models for SFR. As one of the participants, the Nuclear THermal-hydraulic Laboratory at Xi'an Jiaotong University (NuTHeL) participates with the monte carol code OpenMC and system thermal hydraulic code THACS. This paperdeal with the standalone neutronic modeling by OpenMC and thermal hydraulic modelling by THACS. For neutronic part, the axial homogenous model was modeled to analyze the reactivity feedback coefficients and steady radial power distribution. For thermal hydraulic part, the system model, including 2D CFD outlet plenum model, subchannel model and inter-wrapper flow model, was modeled to analyze both system response and local phenomena during the LOSWOS. In general, both neutronic results and thermal hydraulic results agreed well with other participants or experiments, the reactivity feedback of GEM could provide major negative feedback, which could be used as a novel passive shutdown device. However, the remaining discrepancies were the underestimated peak temperature of PIOTA at the begin of transients, it may be caused by unknow heat transfer coefficients around hexagon tube, which has no correlation and should be further investigated by experiments. In addition, when considering IWF model, the maximum temperature of the outlet for the active fuel region was 10 K larger than the subassemblie' outlet in the PIOTA, this phenomenon should be carefully treated in loss of flow accidents of SFR. [ABSTRACT FROM AUTHOR]

The employment of heat pipe reactors is gaining significant traction due to their exceptional advantages in terms of mobility, safety, and applicability across a wide range of scenarios, including strategic defense locations, remote communities, and situations of power shortage. Importantly, most of the heat pipe reactors are designed as solid-state reactors, which avoids the challenges associated with two-phase flow problems that commonly arise from coolant transformation and makes it possible to analyze the reactor through the integration of multiple physical field coupling. In this paper, we conduct a 3D neutronic-thermal-mechanical coupled performance analysis of a heat pipe reactor operating under steady-state conditions. This analysis employs a combination of the Monte Carlo code and finite element analysis software, facilitating in-depth insights into the behavior of reactor. Coupling calculation results show that under normal operation, the maximum temperature and thermal expansion of core is 1338.94 K and 2.53 mm. In the steady-state coupling process, the influence of temperature parameters on neutronic calculation is mainly reflected in k eff and its power level, the maximum temperature drops from 1347.26 K to 1338.94 K, but it has little influence on power distribution. By coupling calculations, the maximum stress of the reactor is reduced by 5.78 MPa. The research results will support the subsequent development of multi-physics coupled transient analysis code for heat pipe reactors. [ABSTRACT FROM AUTHOR]

In this paper, a full scale three-dimensional thermal-hydraulic analysis method for plate-type fuel reactor core is proposed through integrating traditional sub-channel analysis concepts into CFD platform. The coupled heat transfer model between fuel plates and narrow rectangular coolant channels is established. The three-dimensional thermal-hydraulic analysis code CorTAF-PT is developed based on open-source CFD platform OpenFOAM. The CorTAF-PT is validated against with the Qiu-TAKAHASHI experiment and IAEA 10 MW MTR. The code simulation results agree well with the experiment and MTR operation data with relative error within 5%. The full-scale MTR simulation under steady state and LOFA is carried out through channel-level resolution. In partial blockage accident, maximum temperature of coolant, fuel pellet and cladding increase by 9.75 K, 60.65 K and 60.52 K respectively. Furthermore, the double temperature peaks of fuel plates and local coolant reflux with −0.121 m s−1 are revealed under partial blockage accident. This work provides reference value for correlational studies on thermal-hydraulic characteristics of plate-type fuel reactor core. • CorTAF-PT, a full scale three-dimensional thermal-hydraulic analysis code for plate-type fuel reactor core is developed. • CorTAF-PT can efficiently save the computing resource while remaining refined simulation accuracy. • The CorTAF-PT is validated against with the Qiu-TAKAHASHI experiment and IAEA 10 MW MTR. • Transient analysis under partial blockage accident of MTR is carried out while the local coolant reflux is revealed. [ABSTRACT FROM AUTHOR]

Lead-cooled Fast Reactors (LFRs) has become a main candidate choice among the fourth-generation reactors, and Steam Generator Tube Rupture accident (SGTR) is one kind of its important Design Extension Condition (DEC). Due to the limitations of existing theoretical research and simulation tools, many SGTR thermal-hydraulic experiments have been carried out recently. This paper reviews thermal-hydraulics stages into four stages: 1) water-LBE interaction together with steam explosion, 2) high-pressure water jet and structures impact, 3) steam bubbles transport in primary system, 4) bubble entrained and stayed in the reactor core. The existing SGTR experiments are classified as separate SGTR effect test experimental facilities and separate LBE-water interaction experimental facilities. The methods of high-pressure injection, LBE status change and parameters measurement of separate SGTR effect experimental facilities are introduced. Four experimental prospects for further research are suggested as pressure transmission and structure impact, steam lifting and visualization, bubble transport in full-scale or reduced-scale model of reactor, behaviors of LBE-water mixture. • An LFR experimental review of SGTR thermal-hydraulic experiments. • SGTR thermal hydraulic phenomena are divided. • SGTR thermal hydraulic experiments are classified into two categories. • Experimental methods introduction and further research suggestion. [ABSTRACT FROM AUTHOR]

In order to explore the core corrosion characteristics of horizontal lead-bismuth reactor in oxygen-controlled environment, a liquid lead-bismuth eutectic (LBE) corrosion model including the growth and removal process of oxide layer was established in this paper. The coupling method of double-layered oxidation model and computational fluid dynamics (CFD) was used to simulate the long-term oxidation corrosion of horizontal LBE reactor core numerically. The results show that the average total oxide layer thickness of fuel rod surface is 1.59 μm, and the thickest oxide layer is 3.48 μm on the fuel rod surface at the core outlet after 20,000 h. The fuel rods maintain a double-layer oxide layer structure. The growth rate and removal rate of the oxide layer on fuel rod surface increase with the increase of inlet temperature and the decrease of inlet flow rate. The growth rate of the oxide layer on fuel rod surface increases with the increase of inlet oxygen concentration, while the removal rate decreases. At higher temperature above 420℃, lower flow velocity below 0.25 m/s and lower oxygen concentration below 2 × 10−8wt%, the magnetite layer will first be completely removed from the surface of the fuel rods at the core outlet. The liquid LBE oxidation corrosion model and coupled CFD method presented in this paper can be used for numerical simulation of oxidation corrosion in different units of LBE cooling system. [ABSTRACT FROM AUTHOR]

• The Phenix natural circulation test was analyzed by system program THACS. • The predicted results of THACS are in good agreement with the test results. • The effect of Inter-wrapper flow on decay heat removal of pool-type SFR was analyzed. • The analytical capability of THACS program for natural circulation of pool-type SFR was evaluated. For the pool type sodium-cooled fast reactor (SFR), whether the primary circuit can establish a stable natural cycle after the reactor accident, and whether the primary natural cycle can take away the decay heat of the reactor core in the long-term, are the key issues in the design and safety analysis of SFR. In order to verify the natural circulation capacity of the primary circuit, a natural circulation test was carried out on pool type SFR Phenix by CEA before the reactor is decommissioned. One objective of this test is to verify the system codes used to simulate sodium natural convection phenomena. In this paper, the system code THACS is used to analyze Phenix natural circulation test, and the capacity of THACS to analyze the natural circulation of SFR is evaluated. Firstly, this paper introduces the components of Phenix reactor and the process of natural circulation test. Then, the models of program and the computational modeling of Phenix reactor are described. The calculation involves the whole process of the test: dry out of steam generators, scram, stop of pumps, two phases of natural circulation. Then, the results are compared with the test data, and the influence of inter-wrapper flow (IWF) on core outlet temperature is emphasized. The results show good applicability of THACS to analyze natural circulation in SFR, and the IWF can reduce the peak value of core outlet temperature at the beginning of natural cycle. Finally, the deviation between calculation and test results are analyzed, and the limitation of the models are evaluated. [ABSTRACT FROM AUTHOR]

• Code development and analysis of heat pipe cooled PRHRS of MSR are conducted. • Reactor model shows a good agreement with MSRE data and results. • Heat pipe cooled PRHRS model is validated with experimental data of 2.8% maximum error. The high efficient and reliable Passive Residual Heat Removal System (PRHRS) is of great importance to the development of Molten Salt Reactor (MSR), whether the liquid fuel MSR or solid fuel MSR. In this research, based on the structure and operation performance of the new conceptual PRHRS, using high temperature heat pipe as heat transportation, a set of reasonable mathematical and physical models have been built, mainly including reactor core model, high temperature heat pipe model and PRHRS model. The new conceptual Passive Residual heat removal system Analysis Code (PRAC) has been developed using FORTRAN language. The verification and validation for PRAC is conducted by using existed Molten Salt Reactor Experiment (MSRE) data provided by Oak Ridge National Laboratory and the PRHRS experimental data in our previous work. The results show that the developed PRAC could predict the MSR transients reasonably and show high fidelity. The numerical results of reactor core neutronic and thermal–hydraulic model show a good agreement when compared with MSRE pump startup and shutdown transients. The numerical results by PRHRS model agree well with experimental data which the maximum error of the fluoride salt temperature and heat pipe wall temperature are within 1.2% and 2.8%. This research provides a valuable software basis for the design and analysis of PRHRS of MSR. [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]