[en] Based on the four-equation drift flux model, this paper establishes a one-dimension distribution model for the vertical U tube steam generator. The model considers the area of the primary side, the secondary side, the U tube and the steam dome. Firstly, the discrete equations are obtained by using the first order difference method with the staggered grid, and solves with iteration by applying intersected calculation of thermal and hydraulic process. This work is compiled into a simulated program with the MATLAB software. Applying the program to simulate the thermal-hydraulic parameters under steady state operation of Qinshan NPP steam generator (SG), and the calculation is compared with the RELAP5 program. Finally, the operation characteristics of steam generator under 100%, 75%, 50%, 30%, 15% powers are computed and analyzed. (authors)

[en] At present, the numerical solution of one-dimensional two-fluid model is only first order. Safety design of advanced reactor depends on higher-ordered numerical scheme and code. This work aims at the investigation of the second-ordered scheme for the one-dimensional two-fluid model, to apply the second scheme for improving one first-order test code TFIT to second ordered. Flux-limiter scheme is used for the convection term, the surface tension term is adopted to make the numerical solution stable. The classical water wave problem-water faucet problem is used as the benchmark to test the second-ordered scheme. The numerical solutions of second-ordered and first-ordered scheme are compared with analytical solutions, and the comparison shows that the second-ordered solutions tend to analytical solutions better, catch the void wave of the water faucet problem well, and keep to the same stability of first-ordered code. (authors)

[en] In order to investigate the convective heat transfer at supercritical pressures, experimental research was conducted on heated Freon R134a flowing upward through an I.D 25 mm vertical circular tube under supercritical pressure condition. The experimental data covers a wide range of conditions. The pressure is at 4.5 MPa and 4.7 MPa. The mass flow flux varies from 400 to 700 kg/(m²·s) and the heat flux ranges from 30 to 60 kW/m². Both enhanced and deteriorated heat transfer were analyzed and the parametric sensitivity was also carried out. The heat transfer performance was evidently enhanced near the pseudo-critical point. The deteriorated heat transfer appeared at lower mass flow velocity or at higher heat flux under a certain ratio of q/G = 0.06 kj/kg. At a mass flux of 500 kg/(m²·s) two types of deteriorated heat transfer were observed in the experiment: the first type appeared at the near entrance region of the tube and existed within different range of fluid inlet temperature; The second type appeared at any section inside the tube than entrance but only within a certain enthalpy range. Heat transfer can be enhanced by increasing mass flow velocity, decreasing heat flux or decreasing pressure, while the variance of the heat transfer deterioration is opposite. (authors)

[en] To study the flow and heat transfer characteristics of the primary and secondary sides of the helical coil once-through tube steam generator (HCOTSG) under steady state conditions, taking HCOTSG of International Reactor Innovative and Secure (IRIS) as the research object, a primary and secondary sides heat balance calculation model for steady state operation of HCOTSG is established. The influence of different secondary side feed water flow rate on HCOTSG thermal and hydraulic parameters under steady-state condition is analyzed, and the detailed thermal and hydraulic parameters in the helical tube under steady-state condition are calculated by combining the coupled thermal analysis model with the three-dimensional flow and heat exchange calculation of CFX. The relevant thermal and hydraulic parameters along the tube side of HCOTSG during steady-state operation are calculated by the thermal analysis model. The CFX simulation results show that the velocity and temperature distribution of the fluid in the cross section of the helical tube are not uniform. The temperature of the fluid inside the helix is higher than that outside the helix. The velocity of the fluid inside the helix is lower than that outside the helix. The boiling of the fluid inside the helix occurs earlier than that outside the helix. Therefore, this study has a guiding role in the accident analysis for HCOTSG steady-state operation and spiral heat exchange tube. (authors)

[en] Highlights: • A methodology for predicting circumferentially non-uniform heat transfer in subchannel analysis is established. • CFD simulations are performed to generate the necessary data for correlating the non-uniform factor. • The prediction model is validated against a 2 x 2 rod bundle experiment. - Abstract: The non-uniformity of circumferential heat transfer is small in conventional PWR core but becomes significant with tight-lattice core design for advanced reactors. Predicting the circumferentially non-uniform heat transfer behavior can be challenging given the considerable heterogeneity of the subchannel geometry and the drastic change of property with supercritical fluids. In this paper, a circumferentially non-uniform heat transfer model for subchannel analysis has been developed to predict the circumferential distributions of heat transfer coefficient, wall temperature and wall heat flux. In the model, the sources of the heat transfer non-uniformity are considered to be the circumferentially non-uniform flow area and the fluid property variation. To account for these two effects, new correlation with a non-uniform factor is developed. A series of tests using CFD method was performed for determining the empirical coefficients of the non-uniform factor. Furthermore, a two-dimensional fuel heat conduction model is also added to the subchannel analysis code. The new model was validated by comparing the prediction results with available experimental data of a 2 × 2 square rod bundle with supercritical water. It is demonstrated that the inclusion of circumferentially non-uniform heat transfer model leads to an improvement in the predictive capabilities for current subchannel analysis method and will improve the prediction accuracy of cladding temperatures in the design and safety analysis of reactor fuel elements.

[en] Highlights: • Heat transfer experiments on a tight 19 rod bundle have been performed. • The cause of circumferentially non-uniform wall temperature are discussed. • Parametric effects on circumferential non-uniformity are studied. - Abstract: The strong non-uniform distributions of circumferential wall temperature and heat transfer coefficient will exist in tight-lattice rod bundles. However, there is still a lack of understanding and only a few experimental data are available. In this paper, experimental investigations are performed on circumferential heat transfer behavior in a tight hexagonal 19 rod bundle using supercritical R134a as working fluid. The rod surface temperatures at various circumferential positions along the axial flow direction are measured. A defined local hydraulic diameter is introduced to represent the heterogeneity of channels in the tight bundle. The circumferentially non-uniform distribution of wall temperature is found to be strongly related to the variation of the local hydraulic diameter. In normal heat transfer pattern, the circumferential temperature gradient is large especially in the region far away the pseudo critical point, but becomes small near the pseudo critical point due to a strong enhancement of heat transfer coefficient. Parametric effects of pressure, mass flux and heat flux on circumferential non-uniformity of wall temperature inside the rod bundle are also analyzed.

[en] Highlights: • Convenient set of two-fluid equations and its associated solution method are introduced. • All unknown variables can be obtained by a convenient one-step coupled solution. • The model is validated against water faucet problem and subcooled boiling experiment. - Abstract: Most of the system codes for reactor Thermal hydraulics simulation are based on the two-fluid model. The accurate and efficient simulations of the two-fluid model are therefore important. This paper introduces a numerically convenient set of equations and the associated one-step coupled solution method for the two-phase two-fluid model. These equations are devised from the RELAP5 difference equations, but they are cast into a new form having great numerical effectiveness in forming a coupled matrix equation. A one-step solution of the coupled matrix equation is performed to obtain the whole unknown variables of the flow system. Proper numerical treatments, closure relations and coupling of wall conduction are accounted for by modeling boiling phase change convective heat transfer behavior. Finally, the numerical solutions of the new two-fluid model equations were successfully implemented and validated using the water faucet problem and a subcooled flow boiling experiment, and good agreements were obtained.

[en] Critical (choked) flow is a highly concerning phenomenon in safety analysis for nuclear energy. The discharge mass flow rate prediction is crucial for engineering design and emergency response in case of nuclear accidents. Unfortunately, the critical flow is difficult to predict especially when the two-phase flow exists. The accuracy is based on a deeper understanding of the complex phenomenon of critical flow. The influence of virtual mass force on the two-phase critical flow was seldom concentrated on owing to the lack of suitable critical flow models for studies in detail. This study is based on a developed 6-equation two-phase critical flow model. It is confirmed that the virtual mass force contributes to the stability and convergence of the critical flow simulation and it will impact not only the critical mass flux but also the thermal hydraulic parameters along the discharge duct. The magnitude depends on the geometry of the discharge duct and the upstream condition. It is larger when the duct is longer and the pressure is lower. Furthermore, the virtual mass force for each flow regime was studied in detail with a sensitivity study. The results show that the most sensible condition for the virtual mass force is annular flow along a long tube under relatively low pressure. The future work is to develop a correlation of virtual mass force for critical flow specifically since the correlations in the literature were developed under general two-phase flow process conditions.

[en] Highlights: • Thermal-hydraulic analysis model based on two-fluid model is developed for HCOTSG. • Specific correlations for flow and heat transfer in helical tube are adopted. • Governing equations and heat conduction equation are solved simultaneously. • Steady-state and transient behaviors of the HCOTSG are investigated. - Abstract: A thermal-hydraulic analysis model, coupling the flow and heat transfer of the primary and secondary sides, is developed to describe the thermal hydraulic behavior of OTSG (helically coiled Once-Through Steam Generator). This numerical model is developed based on the two-fluid model with distributed parameters method to predict all flow variables at each position along the tube. Correlations, validated a lot, are adopted to consider the effect of helical structure on the flow and heat transfer of both the primary and secondary sides in OTSG. The computational code THOSG (Thermal-Hydraulic analysis code of Once-through Steam Generator) for OTSG is then developed based on the numerical model with the modified SIMPLE algorithm. To benchmark the developed physic model and computer code, steady-state simulation of OTSG of IRIS reactor is conducted. Results are compared with RELAP5 code with respect to the design parameters. Comparisons indicate that the THOSG code can predict the thermal-hydraulic characteristics of OTSG well. In order to investigate the effect of changes in the input conditions on the output of the OTSG, transient analysis has been performed. Simulations of change in feedwater flow rate and temperature are conducted to assess the performance of OTSG. Both steady-state and transient results indicate that the THOSG code can be utilized for the design and performance analysis of OTSG. However, further analysis, coupling with the thermal-hydraulic behavior of the reactor core, is required to comprehensively evaluate the safety and reliability of OTSG design in nuclear power plant.

[en] Highlights: • A one-dimensional two sides heat balance model steady state HCOTSGs is established. • The thermo-hydraulic characteristic parameters are obtained. • A CFD simulation model of HCOTSGs secondary circuit is established. • The phenomenon of secondary flow is analyzed. To examine the thermo-hydraulic characteristics of the primary and secondary sides of a helical tube once-through steam generator (HCOTSG) under steady state conditions, a one-dimensional primary and secondary side heat balance calculation model is established. The accuracy of the numerical simulation is verified by a comparison with the results of a RELAP5 program. The thermo-hydraulic parameters of HCOTSGs are obtained using the calculation program. To study the detailed flow and heat transfer characteristics in helical tubes, a CFD simulation model of HCOTSGs secondary side is established. The velocity and temperature distributions of the fluid in the cross section of the helical tube are nonuniform. The temperature of the fluid inside the helix is higher than that outside the helix. The velocity of the fluid inside the helix is lower than that outside the helix. The fluid inside the helix begins to boil before that outside the helix.