bluuebunny/crossref_metadata_2025_split huggingface dataset:
Research on Characteristics of Condensation Heat Transfer for Marine Nuclear Power Platform PRHR HX Under Tilting Condition
Authors: Li Pengzheng , Li Yongquan , Liu Shaoyou , Zhu Dongbao , Zhu Zhiqiang , Kong Xiaming 
Published: December 2022

Abstract
Passive residual heat removal heat exchanger (PRHR HX) is one of the important equipment in passive residual heat removal system (PRHRs) of marine nuclear power platform. The research on the condensation heat transfer characteristics of saturated steam in PRHR HX tube under tilting condition can provide support and optimization for passive safety system design of marine nuclear power platform. The work in this paper is of great significance to safe operation of marine nuclear power platform. The heat transfer characteristics of saturated steam condensation in the PRHR HX tube under tilting condition are analyzed by building an experimental device with a power ratio of 1:50. The experimental results show that within the range of experimental parameters in this paper, compared with the static state of the experimental device, the condensation heat transfer coefficient of saturated steam in PRHR HX tube under tilting condition is increased. When the heat flux is 190kW/m2, the heat transfer coefficient of saturated steam condensation in PRHR HX tube under tilting condition increases by about 40% compared with static condition. In this paper the formula for calculating the condensation heat transfer coefficient of saturated steam in PRHR HX tube is revised by introducing the tilt angle. The relative error between the modified formula and the experimental value is within ±10%. The research results of this paper can provide reference for the design and optimization of passive safety systems for marine nuclear power platform and similar applications.
One Dimensional Thermal Steady State Code of Sodium Heated Large Straight Tube Steam Generator
Authors: Yang Dingbang , Zhu Lina , Yu Huajin , Yan Xiangyu 
Published: December 2022

Abstract
Sodium heated once through steam generator (SG) is an important part of sodium cooled fast reactor. It converts the heat of hot sodium into water and water into superheated steam. Steam can turn a turbine to generate electricity. Compared with modular steam generator, large steam generator has great advantages in the application of large commercial sodium cooled fast reactor. Large steam generator can reduce its manufacturing cost and improve the economy of sodium cooled fast reactor. At the same time, the design of double-wall tube structure reduces the probability of sodium-water reaction and improves the safety and reliability of steam generator. In this paper, the heat transfer regions of the fluid in the heat transfer tube were reasonably divided, the heat transfer relations suitable for the double-wall tube steam generator were compared and analyzed, the contact thermal resistance of the double-wall tube was studied, and a set of physical models suitable for One-dimensional Steady-state calculation of large double-wall straight tube steam generator were established. In order to test the established physical model, the calculation results of the program were compared with the design data of some double-wall straight tube steam generators. It was found that the two agree well, which verified the correctness of the program. The 1-D steady-state thermal analysis of double-wall straight tube steam generator (DWTSG) was carried out by using the calculated results of the code.
Numerical Study on Fluctuation Characteristics of Free Liquid Surface Under External Force
Authors: Wei Xiaojing , Xia Genglei 
Published: December 2022

Abstract
Ocean conditions can cause the steam generator in nuclear-powered ship to heave, sway and tilt with the ship, causing the free liquid surface in the steam generator to fluctuate and causing additional fluctuations of pressure and water level. In this paper, the CFD method is used to numerically study the fluctuation characteristics of the free liquid surface during the sloshing process of the secondary side liquid in the steam generator under the action of external force. The free liquid surface is tracked by using the VOF (Volume of Fluid Method) method, and additional inertial force is added through the UDF (User Define Function) to simulate rolling conditions. First, the simulation results of the numerical model in this study are compared with the experimental results, and the results are in good agreement, which verifies the reliability of the numerical model to analyze the liquid sloshing problem. Then, through numerical simulation, the fluctuation characteristics of the free liquid surface and the change of the water level and the pressure in the steam generator under different excitation periods are obtained. The results show that the liquid sloshing process in the steam generator exhibits periodic changes, and due to the complexity of the structure of the steam generator, the instability and nonlinearity of the liquid sloshing are enhanced.
Adaptability Analysis of Resistance and Heat Transfer Characteristics of Different Turbulence Models in Narrow Rectangular Channel
Authors: Yu Shengzhi , Gong Zili , Jia Zhen , Kuang Zhengfei , Yuan Anming , Zhou Ping , Yang Xu 
Published: December 2022

Abstract
Under ocean conditions, the flow of natural circulation system will fluctuate periodically, and the flow state of single-phase fluid flowing through the heating channel will change in real time, which may be one or more states of laminar flow, transition flow and turbulence. At present, the main CFD calculation programs, in order to be able to solve quickly, the solution of laminar flow process and turbulent process are basically separated. It is difficult to solve the fluid motion process with multiple flow states with the same set of models, which makes it difficult to study the characteristics of flow and heat transfer in this state. Firstly, the pressure drop experimental results in the channel are used to verify and analyze the flow and heat transfer capacity of eight turbulence models in the narrow rectangular channel, and a turbulence model suitable for analyzing the flow and heat transfer characteristics when multiple flow states coexist or change periodically in the narrow rectangular channel is found. At the same time, the model results are verified by multiple groups of experimental results of resistance and heat transfer characteristics, It provides an idea for the prediction and analysis of flow field and heat transfer characteristics under ocean conditions.
Development and Validation of Simulation Method for a Two-Phase Flow Ejector
Authors: Nakanishi Hiroaki , Komuro Yoshiteru , Kondo Yoshiyuki , Tanimoto Koichi 
Published: December 2022

Abstract
A two-phase ejector is a device to induce a suction flow without pump or electricity. The flow in the two-phase ejector consists of a drive flow and a suction flow. As the driving flow expands blowing out of a drive flow nozzle, the thermal energy potential is converted into momentum, and by giving it to the suction flow, it is possible to induce the flow without using external power. In a nuclear power plant, a two-phase ejector can be utilized as a device to drive coolant flow in the cases of power failure.
Mixing of the drive flow and the suction flow accompanied with evaporation or condensation at the gas-liquid interface depends on thermal hydraulic parameters and flow rate, and it is necessary to control them to maintain the driving force, but it can easily come out of operation range with a slight change in balance. There is little knowledge about heat and mass transfer to find and design operating conditions and ejector configurations.
In this study, a heat and mass transfer model of the gas-liquid interface in a critical two-phase flow was developed. To handle thermally non-equilibrium two-phase flow with phase changes occurring simultaneously at the interface, we implemented constitutive equations into CFD tool, such as a correlation for interfacial area concentrations, and we evaluated evaporation coefficient, which is an important parameter to determine the phase change rate, based on the physical property of the working fluid. The CFD simulation method was validated using the experimental data in the literature of a two-phase ejector. In the validation, the flow rates of the drive flow and the suction flow, and pressure distribution inside the ejector were compared. Then, the validity of the developed CFD simulation method have confirmed.
Calculation Method and Analysis on Thermal Compressor Ejection Characteristics
Authors: Xu Huiqiang , Dai Shoubao , Sun Na , Sun Lin , Wang Yanqing 
Published: December 2022

Abstract
Thermal compressor, a similar device like steam driven jet pump, is usually used to keep the vacuum level of condensers in nuclear power plant. Meanwhile, the utilization of thermal compressor is an effective way to recover the heat of exhausted steam from turbines, improving the energy efficiency of secondary circuit in nuclear power plants. The extreme entrainment ration of thermal compressor is a significant parameter to describe its working performance. The present research conducts the theoretical derivation of the related physical process inside the thermal compressor and builds a theoretical model calculating the maximum entrainment ration. The results show that the present model can predict the ejection ability of the thermal compressor with satisfactory accuracy and application scope. The average relative deviation between calculated and experimental result is 9.7% and the maximum deviation does not exceed 23%. Moreover, the results of the model calculation indicate that the ejection ability is enhanced with the increase of primary and suction steam pressure, but weakened by the increase of compressed steam pressure. The entrainment ration increases with the superheat degree of primary steam because of the increasing critical velocity while the superheat degree of suction steam makes no obvious influence.
Development and Verification of Thermal-Hydraulic Constitutive Model for Rectangular Channel
Authors: Feng Yun Fei , Li Lei , Nie Ya Jie , Jiao Xin , Li Jian Yi , Meng Si Jia , Li Ying Hao 
Published: December 2022

Abstract
The plate-shaped fuel element has good heat transfer characteristics, high average power density of the core, and low temperature of the fuel core, which is beneficial to improve the power-to-volume ratio of the core and ensure the safety of the core. Therefore, plate fuels are widely used in compact reactors such as research reactors, integrated reactors, and high-flux reactors. at present, most thermal-hydraulic analysis programs, such as RELAP, RETRAN, THEATRE, are mostly developed for large-scale pressurized water reactors using rod-shaped fuels. It is suitable for narrow rectangular channel of plate type fuel core. Based on this, this paper developed a set of thermal-hydraulic constitutive relation models suitable for narrow rectangular channels, including: flow resistance coefficient calculation model, wall heat transfer Coefficient calculation model, CHF calculation model, etc. The thermal-hydraulic constitutive relational model library of rectangular channel of plate-shaped fuel element is developed by using C++ language. In this paper, the developed constitutive relation model is transplanted into the reactor thermal-hydraulic real-time simulation program, and the IAEA 10MW material test reactor (MTR) benchmark is used to verify the developed rectangular channel thermal-hydraulic constitutive relation model library. Simulation analysis is carried out for two typical accident conditions, reactive introduction (RIA) and loss of flow accident (LOFA) defined in the benchmark problem. correctness.
Study on Fluid-Structure Interaction of Fuel Rod Bundle With Spacer
Authors: Li Shuang , Sun Lanxin 
Published: December 2022

Abstract
In the reactor, the axial flow of the coolant into the core causes the rod bundle to vibrate, and the long-term flow-induced vibration will cause fretting wear. In severe cases, the fuel cladding will be destroyed, jeopardizing the efficiency and safety of the reactor. Therefore, the study of the fluid-induced fluid-structure interaction of fuel bundles is beneficial to ensure the safety of the reactor. In this paper, based on computational fluid dynamics and computational structural dynamics, using Euler-Bernoulli beam theory (Euler-Bernoulli), the numerical calculation of fluid-structure interaction is realized by Fluent. And compared with the 5 × 5 fuel bundle without grid.The influences of the rod bundles with or without grids on the fluid pressure, fluid displacement and vibration under different cross-sections are mainly analyzed, as well as the principle of flow-induced vibration in axial flow. The results show that the influence of the vibration of the rod bundle on the flow field is very weak and can be ignored; The pressure difference between the two radial ends of the fuel rod bundle is one of the factors that cause the flow-induced vibration; The existence of the spacer grid aggravates the disturbance of the flow field, so that a larger pressure difference between the two rod bundles causes a larger vibration, and also affects the position where the larger vibration displacement occurs.
Three-Dimensional Fine-Mesh Coupled Neutronics and Thermal-Hydraulics Calculation for PWR Fuel Pins
Authors: Xie Qiuxia , Liu Xiaojing , Chai Xiang 
Published: December 2022

Abstract
With the great improvement of computer performance, multi-physics and high-fidelity reactor numerical simulation has attracted widespread attention. The development of a three-dimensional fine-mesh coupled neutronics and thermal-hydraulics procedure mainly using the free open source C++ library OpenFOAM is presented in this research. The coupling codes are allowed to solve the coupled neutronic and thermal-hydraulic problem within the same one procedure for both steady-state and transient conditions, avoiding the data transmission between different programs. The coupling strategy among two physical fields is implemented in the same fully three-dimensional and fine mesh system, which eliminates numerical errors caused by the mapping of the grid. For the thermal-hydraulics, the built-in solid-fluid coupling conjugate heat transfer solver is applied for the calculation of the fluid mass, momentum, and energy equations, together with the solid energy equation to obtain the temperature distribution. The neutron diffusion equation is solved iteratively via the developed three-dimensional multi-group multi-region neutron diffusion solver to get the power distribution. The macroscopic cross-sections are pre-generated by the Monte Carlo code OpenMC and fitted as functions of temperature added in the neutron diffusion solver. The temperature distribution obtained by thermal-hydraulics calculation will change the macroscopic cross-sections and then impact the neutron diffusion calculation, while the power distribution gained from the neutronic calculation is transferred to the thermal-hydraulics calculation and is used as the heat source term. This coupling methodology are tested on a 3 × 3 PWR fuel pins model. The results show that all physical fields conform correct distribution regularity, illustrating the feasibility of the coupling methodology.
Analysis of the Thermal Performance of the Passive Containment Cooling System Based on Heat Pipe Technology Under Nuclear Accidents
Authors: Wenying Ji , Changdong Lu , Yangyong Ou , Yichen Chen 
Published: December 2022

Abstract
To evaluate the thermal performance of a closed-loop Passive Containment Cooling System (PCCS) based on hot pipe technology under nuclear accidents, the coupled effect of three systems of reactor primary cooling circuit, containment and the PCCS is analyzed under LB-LOCA with SBO and SB-LOCA with SBO with the PCCS started one hour after the accident. The coupled analysis method is that the heat transfer power of PCCS simulated by the LOCUST model and the mass and energy releasing results obtained from the integrated analysis program for serious accidents are the inputs for containment thermal response analysis code. The results showed that in the 0∼10 hours after the accident, the containment pressure is less than the design limit of 5.20bar.a; in the middle and long term of ten more hours after the accident, the containment pressure is less than 3.50bar.a, the integrity of the containment can be ensured and the feasibility of the design of the PCCS in this paper is validated.
Assessment on Partial Derivatives for Thermal-Physical Properties of Carbon Dioxide
Authors: Wen Shuang , Wen Qinglong 
Published: December 2022

Abstract
To adapt to a requirement of improving the accuracy and efficiency of calculation, a full or partial implicit scheme is usually employed in solving the conservative equations of the supercritical carbon dioxide (S-CO2) Brayton cycle, and partial derivatives of thermal properties such as (∂h/∂ρ)p and (∂h/∂p)ρ are needed in numerical solver. In this paper, the most representative state equations of carbon dioxide are investigated and evaluated by experimental data. The Span-Wagner (SW) equation has a minimal error in all state equations, so the SW equation is chosen as the fundamental equation of thermal properties for partial derivatives. Based on that, the equations of partial derivatives such as (∂h/∂ρ)p and (∂h/∂p)ρ are presented by the Maxwell equation. The paper also evaluates the closure of partial derivatives equations. The deviations of (∂h/∂ρ)p and (∂h/∂p)ρ are within ±0.01% for most points. The maximum closure error of (∂h/∂ρ)p is 0.373%, and the maximum one of (∂h/∂p)ρ is −0.798%. Therefore, the partial derivatives equations obtained in this paper can play a significant role in the safety analysis code.
Research on the Effect of Non-Uniform Inflow on the Cavitation Characteristics of the RCP
Authors: Song Yu , Huang Song , Xu Rui , Yin Junlian , Wang Dezhong 
Published: December 2022

Abstract
In Advanced Passive (AP) series reactors, the reactor coolant pump (RCP) is directly connected to the bottom of the steam generator (SG) rather than a longer pipe for the fully development of inflow, which leads to an unsteady flow at pump inlet. Some scholars have done several study on the cavitation characteristics of RCP. However, the reality is that two pumps are operating at the same time with different inflow. There has been little research about this. Thus, in this paper, a special test loop has been established to simulate this situation. Then numerical and experimental study were carried out with the special system. As a result, cavitation characteristics of parallel pumps were illustrated under the condition of non-uniform inflow, and these results were compared with the uniform condition. The results show some differences in cavitation performance between different conditions. On the one hand the non-uniform incoming flow causes the angle of attack of the impeller changed, and the distribution of cavitation on the blades varies greatly. On the other hand, there are also some difference in cavitation characteristic between the two pumps duo to the distinction of inflow. The cavitation performance has a little change in rated and small flow condition, but for large flow condition has been a dramatic deterioration.
Numerical Simulation of Transient Thermal-Hydraulic Characteristics of Intermediate Heat Exchanger for Sodium Cooled Fast Reactor Under Accident Conditions
Authors: Zeng Xiehu , Wen Qinglong , Bai Genxing 
Published: December 2022

Abstract
In Generation IV nuclear systems, Sodium Cooled Fast Reactor has attracted extensive attention for its particular advantages. Intermediate Heat Exchanger (IHX) is a significant equipment connecting the primary circuit system and secondary circuit system in Sodium Cooled Fast Reactor. Under normal and accident conditions, the temperature fluctuation of the IHX component may cause the alternating change of the local stress, such as sealing weld zones mixing chamber, and the area between inner and outer sleeves. This fluctuation may result in the thermal fatigue of IHX and consequently affect the safety and economy of the reactor operation. Therefore, it is essential to carry out mechanical analysis by experiments or simulations to ensure the structural stability under complex conditions. However, the mechanical analysis must take transient thermal-hydraulic characteristics as boundary and input conditions. Thus, a thorough thermal-hydraulic assessment of IHX is required to guarantee its security under accident conditions.
In this paper, a thermal-hydraulic simulation of IHX was carried out using system code under steady and transient state conditions. Thermal parameters of steady-state calculation agreed well with the design requirements. Transient-state accident conditions, such as emergency shutdown, Station Blackout (SBO), and Steam Generator Tube Rupture (SGTR), were conducted in this paper. In the emergency shutdown, the wall temperature increased from 703.15 K to 798.41 K in three seconds, and then decreased slowly and stabilized at 626.15K. In the SBO accident, the temperature of the primary and secondary circuit fluids fluctuated violently from 0 to 100 seconds. When the secondary side flow drops to 0, the wall temperature of typical positions changes with the inlet temperature of the primary side, showing a trend of rapid decline and stable. In the SGTR accident, the temperature of the heat transfer tube wall increased rapidly during the early stage of the accident. After that, the primary side flow reduced gradually to 0 with the decrease of the secondary side flow. Because of the trend of flow, the temperature of the tube wall decreased rapidly and then increased slowly. Therefore, the results of various accident conditions in this investigation can contribute to the thermal fatigue analysis of IHX in the near future.
Investigation on a Complete Passive Cooling System Using Large-Scale Separate Heat Pipes in Spent Fuel Pool
Authors: Han Fei , Chen Xiting , Kuang Yiwu , Wang Wen , Ye Cheng 
Published: December 2022

Abstract
Large-scale separate heat pipes used in the complete passive cooling system (PCS) transfer the decay heat in the spent fuel pool (SFP) efficiently through the two-phase natural circulation without any external power. In this study, a lumped mathematical model for the heat pipes are developed and parameters related to the heat transfer ability are discussed to settle the number of the heat pipes under different heat load. For the condensers at the auxiliary building, the effects of the tube pitch and the fin height are discussed, which are key parameters to the heat transfer performance. Different structural designs of the PCS under typical operating conditions are settled. A larger quantity of heat pipes is required for higher decay heat power conditions. To validate the reliability of the PCS, transient three-dimensional simulations of the SFP with immersed evaporators under different heat loads are conducted. Based on the results, detailed thermal-hydraulic characteristics are captured in the pool. Large natural convection circulations are formed at the steady-state. Single flow circulation is formed in the X-Z plane under low heat load cases while a pair of counter-rotate natural convection circulations under high heat load cases. A larger heat load promotes the natural convection intensity and shortens the response time of the PCS. Proper distance between the heat source and heat sink in both vertical and horizontal directions in the SFP is beneficial to the flow organization, improving the heat transfer efficiency of the PCS. The maximum temperature in the SFP is always below the saturation point after the startup of the heat pipes, which could validate the reliability of the PCS and ensure the safety of the plant under emergency conditions.