Performance analysis of automatic depressurization system in advanced PWR during a typical SBLOCA transient using MIDAC

Sun, Hongping,Zhang, Yapei,Tian, Wenxi,Qiu, Suizheng,Su, Guanghui Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.5

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

Numerical study of laminar flow and friction characteristics in narrow channels under rolling conditions using MPS method

Muhammad Abdul Basit,Wenxi Tian,Ronghua Chen,Suizheng Qiu,Guanghui Su 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.8

Modern small modular nuclear reactors can be built on a barge in ocean, therefore, their flow charac-teristics depend upon the ocean motions. In the present research, effect of rolling motion on flow and friction characteristics of laminar flow through vertical and horizontal narrow channels has been studied. A computer code has been developed using MPS method for two-dimensional Navier-Stokes equations with rolling motion force incorporated. Numerical results have been validated with the literature and have been found in good agreement. It has been found that the impact of rolling motions on flow characteristics weakens with increase in flow rate and fluid viscosity. For vertical narrow channels, the time averaged friction coefficient for vertical channels differed from steady friction coefficient. Furthermore, increasing the horizontal distance from rolling pivot enhanced the flow fluctuations but these stayed relatively unaffected by change in vertical distance of channel from the rolling axis. For horizontal narrow channels, the flow fluctuations were found to be sinusoidal in nature and their magnitude was found to be dependent mainly upon gravity fluctuations caused by rolling.

The development of high fidelity Steam Generator three dimensional thermal hydraulic coupling code: STAF-CT

Zhao, Xiaohan,Wang, Mingjun,Wu, Ge,Zhang, Jing,Tian, Wenxi,Qiu, Suizheng,Su, G.H. Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.3

The thermal hydraulic performances of Steam Generator (SG) under both steady and transient operation conditions are of great importance for the safety and economy in nuclear power plants. In this paper, based on our self-developed SG thermal hydraulic analysis code STAF (Steam-generator Thermalhydraulic Analysis code based on Fluent), an improved new version STAF-CT (fully Coupling and Transient) is developed and introduced. Compared with original STAF, the new version code STAF-CT has two main functional improvements including "Transient" and "Fully Three Dimensional Coupling" features. In STAF-CT, a three dimensional energy transferring module is established which can achieve energy exchange computing function at the corresponding position between two sides of SG. The STAF-CT is validated against the international benchmark experiment data and the results show great agreement. Then the U-shaped SG in AP1000 nuclear power plant is modeled and simulated using STAF-CT. The results show that three dimensional flow fields in the primary side make significant effect on the energy source distribution between two sides. The development of code STAF-CT in this paper can provide an effective method for further SG high fidelity research in the nuclear reactor system.

CFD/RELAP5 Coupling Analysis of the ISP No. 43 Boron Dilution Experiment

Linrong Ye,Hao Yu,Mingjun Wang,Qianglong Wang,Wenxi Tian,Suizheng Qiu,G.H. Su 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.1

Multi-dimensional coupling analysis is a research hot spot in nuclear reactor thermal hydraulic study and both the full-scale system transient response and local key three-dimensional thermal hydraulic phenomenon could be obtained simultaneously, which can achieve the balance between efficiency and accuracy in the numerical simulation of nuclear reactor. A one-dimensional to three-dimensional (1D-3D) coupling platform for the nuclear reactor multi-dimensional analysis is developed by XJTU-NuTheL (Nuclear Thermal-hydraulic Laboratory at Xi'an Jiaotong University) based on the CFD code Fluent and system code RELAP5 through the Dynamic Link Library (DLL) technology and Fluent user-defined functions (UDF). In this paper, the International Standard Problem (ISP) No. 43 is selected as the benchmark and the rapid boron dilution transient in the nuclear reactor is studied with the coupling code. The code validation is conducted first and the numerical simulation results show good agreement with the experimental data. The three-dimensional flow and temperature fields in the downcomer are analyzed in detail during the transient scenarios. The strong reverse flow is observed beneath the inlet cold leg, causing the de-borated water slug to mainly diffuse in the circumferential direction. The deviations between the experimental data and the transients predicted by the coupling code are also discussed.

Numerical analysis of melt migration and solidification behavior in LBR severe accident with MPS method

Jinshun Wang,Qinghang Cai,Ronghua Chen,Xinkun Xiao,Yonglin Li,Wenxi Tian,Suizheng Qiu,G.H. Su 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.1

In Lead-based reactor (LBR) severe accident, the meltdown and migration inside the reactor core will lead to fuel fragment concentration, which may further cause re-criticality and even core disintegration. Accurately predicting the migration and solidification behavior of melt in LBR severe accidents is of prime importance for safety analysis of LBR. In this study, the Moving Particle Semi-implicit (MPS) method is validated and used to simulate the migration and solidification behavior. Two main surface tension models are validated and compared. Meanwhile, the MPS method is validated by the l-plate solidification test. Based on the improved MPS method, the migration and solidification behavior of melt in LBR severe accident was studied furthermore. In the Pb–Bi coolant, the melt flows upward due to density difference. The migration and solidification behavior are greatly affected by the surface tension and viscous resistance varying with enthalpy. The whole movement process can be divided into three stages depending on the change in velocity. The heat transfer of core melt is determined jointly by two heat transfer modes: flow heat transfer and solid conductivity. Generally, the research results indicate that the MPS method has unique advantage in studying the migration and solidification behavior in LBR severe accident.

Experimental study on the Influence of Heating Surface Inclination Angle on Heat Transfer and CHF performance for Pool Boiling

Chenglong Wang,Panxiao Li,Dalin Zhang,Wenxi Tian,Suizheng Qiu,G.H. Su,Jian Deng 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.1

Pool boiling heat transfer is widely applied in nuclear engineering fields. The influence of heating surface orientation on the pool boiling heat transfer has received extensive attention. In this study, the heating surface with different roughness was adopted to conduct pool boiling experiments at different inclination angles. Based on the boiling curves and bubble images, the effects of inclination angle on the pool boiling heat transfer and critical heat flux were analyzed. When the inclination angle was bigger than 90°, the bubble size increased with the increase of inclination angle. Both the bubble departure frequency and critical heat flux decreased as the inclination angle increased. The existing theoretical models about pool boiling heat transfer and critical heat flux were compared. From the perspective of bubble agitation model and Hot/Dry spot model, the experimental phenomena could be explained reasonably. The enlargement of bubble not only could enhance the agitation of nearby liquid but also would cause the bubble to stay longer on the heating surface. Consequently, the effect of inclination angle on the pool boiling heat transfer was not conspicuous. With the increase of inclination angle, the rewetting of heating surface became much more difficult. It has negative effect on the critical heat flux. This work provides experimental data basis for heat transfer and CHF performance of pool boiling.

Molecular dynamics study of liquid sodium film evaporation and condensation by Lennard-Jones potential

Wang Zetao,Guo Kailun,Wang Chenglong,Zhang Dalin,Tian Wenxi,Qiu Suizheng,Su Guanghui 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.8

Deeply understanding the phase change of thin liquid sodium film inside wick pore is very important for further studying high-temperature sodium heat pipe's heat transfer. For the first time, the evaporation and condensation of thin liquid sodium film are investigated by the Lennard-Jones potential of molecular dynamics. Based on the startup and normal operation of the sodium heat pipe, three different cases are simulated. First, the equilibrium is achieved and the Mass Accommodation Coefficients of the three cases are 0.3886, 0.2119, 0.2615 respectively. Secondly, the non-equilibrium is built. The change of liquid film thickness, the number of gas atoms, the net evaporation flux (Jnet), the heat transfer coefficient (h) at the liquid-gas interface are acquired. Results indicate that the magnitude of the Jnet and the h increase with the basic equilibrium temperature. In 520e600 K (the startup of the heat pipe), the h has approached 5 e6Wm2 K1 while liquid film thickness is in 11e13 nm. The fact shows that during the initial startup of the sodium heat pipe, the thermal resistance at the liquid-gas interface can't be negligible. This work is the complement and extension for macroscopic investigation of heat transfer inside sodium heat pipe. It can provide a reference for further numerical simulation and optimal design of the sodium heat pipe in the future.

Water film covering characteristic on horizontal fuel rod under impinging cooling condition

Zhang Penghui,Wang Bowei,Chen Ronghua,Su G.H.,Tian Wenxi,Qiu Suizheng 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.11

Jet impinging device is designed for decay heat removal on horizontal fuel rods in a low temperature heating reactor. An experimental system with a fuel rod simulator is established and experiments are performed to evaluate water film covering capacity, within 0.0287e0.0444 kg/ms mass flow rate, 0 e164.1 kW/m2 heating flux and 13.8e91.4C feeding water temperature. An effective method to obtain the film coverage rate by infrared equipment is proposed. Water film flowing patterns are recoded and the film coverage rates at different circumference angles are measured. It is found the film coverage rate decreases with heating flux during single-phase convection, while increases after onset of nucleate boiling. Besides, film coverage rate is found affected by Marangoni effect and film accelerating effect, and surface wetting is significantly facilitated by bubble behavior. Based on the observed phenomenon and physical mechanism, dry-out depth and initial dry-out rate are proposed to evaluate film covering potential on a heating surface. A model to predict film coverage rate is proposed based on the data. The findings would have reliable guide and important implications for further evaluation and design of decay heat removal system of new reactors, and could be helpful for passive containment cooling research.

Assessment of the severe accident code MIDAC based on FROMA, QUENCH-06&16 experiments

Shihao Wu,Yapei Zhang,Dong Wang,Wenxi Tian,Suizheng Qiu,G.H. Su 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.2

In order to meet the needs of domestic reactor severe accident analysis program, a MIDAC (Module Invessel Degraded severe accident Analysis Code) is developed and maintained by Xi'an Jiaotong University. As the accuracy of the calculation results of the analysis program is of great significance for theformulation of severe accident mitigation measures, the article select three experiments to evaluate theupdated severe accident models of MIDAC. Among them, QUENCH-06 is the international standardNo.45, QUENCH-16 is a test for the analysis of air oxidation, and FROMA is an out-of-pile fuel rod meltingexperiment recently carried out by Xi'an Jiaotong University. The heating and melting model withlumped parameter method and the steam oxidation model with Cathcart-Pawel and Volchek-Zvonarevcorrelations combination in MIDAC could better meet the needs of severe accident analysis. Although theinfluence of nitrogen still need to be further improved, the air oxidation model with NUREG still has theability to provide guiding significance for engineering practice

SEINA: A two-dimensional steam explosion integrated analysis code

Wu Liangpeng,Sun Ruiyu,Chen Ronghua,Tian Wenxi,Qiu Suizheng,Su G.H. 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.10

In the event of a severe accident, the reactor core may melt due to insufficient cooling. the hightemperature core melt will have a strong interaction (FCI) with the coolant, which may lead to steam explosion. Steam explosion would pose a serious threat to the safety of the reactors. Therefore, the study of steam explosion is of great significance to the assessment of severe accidents in nuclear reactors. This research focuses on the development of a two-dimensional steam explosion integrated analysis code called SEINA. Based on the semi-implicit Euler scheme, the three-phase field was considered in this code. Besides, the influence of evaporation drag of melt and the influence of solidified shell during the process of melt droplet fragmentation were also considered. The code was simulated and validated by FARO L-14 and KROTOS KS-2 experiments. The calculation results of SEINA code are in good agreement with the experimental results, and the results show that if the effects of evaporation drag and melt solidification shell are considered, the FCI process can be described more accurately. Therefore, it is proved that SEINA has the potential to be a powerful and effective tool for the analysis of steam explosions in nuclear reactors.