Program development and preliminary CHF characteristics analysis for natural circulation loop under moving condition

Gui, Minyang,Tian, Wenxi,Wu, Di,Chen, Ronghua,Su, G.H.,Qiu, Suizheng Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.2

Critical heat flux (CHF) has traditionally been evaluated using look-up tables or empirical correlations for nuclear power plants. However, under complex moving condition, it is necessary to reconsider the CHF characteristics since the conventional CHF prediction methods would no longer be applicable. In this paper, the additional forces caused by motions have been added to the annular film dryout (AFD) mechanistic model to investigate the effect of moving condition on CHF. Moreover, a theoretical model of the natural circulation loop with additional forces is established to reflect the natural circulation characteristics of the loop system. By coupling the system loop with the AFD mechanistic model, a CHF prediction program called NACOM for natural circulation loop under moving condition is developed. The effects of three operating conditions, namely stationary, inclination and rolling, on the CHF of the loop are then analyzed. It can be clearly seen that the moving condition has an adverse effect on the CHF in the natural circulation system. For the calculation parameters in this paper, the CHF can be reduced by 25% compared with the static value, which indicates that it is important to consider the effects of moving condition to retain adequate safety margin in subsequent thermal-hydraulic designs.

Prediction of dryout-type CHF for rod bundle in natural circulation loop under motion condition

Huang, Siyang,Tian, Wenxi,Wang, Xiaoyang,Chen, Ronghua,Yue, Nina,Xi, Mengmeng,Su, G.H.,Qiu, Suizheng Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.4

In nuclear engineering, the occurrence of critical heat flux (CHF) is complicated for rod bundle, and it is much more difficult to predict the CHF when it is in natural circulation under motion condition. In this paper, the dryout-type CHF is investigated for the rod bundle in a natural circulation loop under rolling motion condition based on the coupled analysis of subchannel method, a one-dimensional system analysis method and a CHF mechanism model, namely the three-fluid model for annular flow. In order to consider the rolling effect of the natural circulation loop, the subchannel model is connected to the one-dimensional system code at the inlet and outlet of the rod bundle. The subchannel analysis provides the local thermal hydraulic parameters as input for the CHF mechanism model to calculate the occurrence of CHF. The rolling motion is modeled by additional motion forces in the momentum equation. First, the calculation methods of the natural circulation and CHF are validated by a published natural circulation experiment data and a CHF empirical correlation, respectively. Then, the CHF of the rod bundle in a natural circulation loop under both the stationary and rolling motion condition is predicted and analyzed. According to the calculation results, CHF under stationary condition is smaller than that under rolling motion condition. Besides, the CHF decreases with the increase of the rolling period and angular acceleration amplitude within the range of inlet subcooling and mass flux adopted in the current research. This paper can provide useful information for the prediction of CHF in natural circulation under motion condition, which is important for the nuclear reactor design improvement and safety analysis.

Parallelization and application of SACOS for whole core thermal-hydraulic analysis

Minyang Gui,Wenxi Tian,Di Wu,Ronghua Chen,Mingjun Wang,G.H. Su 한국원자력학회 2021 Nuclear Engineering and Technology Vol.53 No.12

SACOS series of subchannel analysis codes have been developed by XJTU-NuTheL for many years and arebeing used for the thermal-hydraulic safety analysis of various reactor cores. To achieve fine whole corepin-level analysis, the input preprocessing and parallel capabilities of the code have been developed inthis study. Preprocessing is suitable for modeling rectangular and hexagonal assemblies with less errorprone input; parallelization is established based on the domain decomposition method with the hybridof MPI and OpenMP. For domain decomposition, a more flexible method has been proposed which candetermine the appropriate task division of the core domain according to the number of processors of theserver. By performing the calculation time evaluation for the several PWR assembly problems, the codeparallelization has been successfully verified with different number of processors. Subsequent analysisresults for rectangular- and hexagonal-assembly core imply that the code can be used to model andperform pin-level core safety analysis with acceptable computational efficiency

Investigation of single bubble behavior under rolling motions using multiphase MPS method on GPU

Basit, Muhammad Abdul,Tian, Wenxi,Chen, Ronghua,Basit, Romana,Qiu, Suizheng,Su, Guanghui Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.6

Study of single bubble behavior under rolling motions can prove useful for fundamental understanding of flow field inside the modern small modular nuclear reactors. The objective of the present study is to simulate the influence of rolling conditions on single rising bubble in a liquid using multiphase Moving Particle Semi-implicit (MPS) method. Rolling force term was added to 2D Navier-Stokes equations and a computer program was written using C language employing OpenACC to port the code to GPU. Computational results obtained were found to be in good agreement with the results available in literature. The impact of rolling parameters on trajectory and velocity of the rising bubble has been studied. It has been found that bubble rise velocity increases with rolling amplitude due to modification of flow field around the bubble. It has also been concluded that the oscillations of free surface, caused by rolling, influence the bubble trajectory. Furthermore, it has been discovered that smaller vessel width reduces the impact of rolling motions on the rising bubble. The effect of liquid viscosity on bubble rising under rolling was also investigated and it was found that effects of rolling became more pronounced with the increase of liquid viscosity.

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.

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.

Thermal-hydraulic analysis of He-Xe gas mixture in 2×2 rod bundle wrapped with helical wires

Chenglong Wang,Siyuan Chen,Wenxi Tian,G.H. Su,Suizheng Qiu Korean Nuclear Society 2023 Nuclear Engineering and Technology Vol.55 No.7

Gas-cooled space reactor, which adopts He-Xe gas mixture as working fluid, is a better choice for megawatt power generation. In this paper, thermal-hydraulic characteristics of He-Xe gas mixture in 2×2 rod bundle wrapped with helical wires is numerically investigated. The velocity, pressure and temperature distribution of the coolant are obtained and analyzed. The results show that the existence of helical wires forms the vortexes and changes the velocity and temperature distribution. Hot spots are found at the contact corners between helical wires and fuel rods. The highest temperature of the hot spots reach 1600K, while the mainstream temperature is less than 400K. The helical wire structure increases the friction pressure drop by 20%-50%. The effect extent varies with the pitch and the number of helical wires. The helical wire structure leads to the reduction of Nusselt number. Comparing thermal-hydraulic performance ratios (THPR) of different structures, the THPR values are all less than 1. It means that gas-cooled space reactor adopting helical wires could not strengthen the core heat removal performance. This work provides the thermal-hydraulic design basis for He-Xe gas cooled space nuclear reactor.

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

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.

Numerical simulation of natural convection around the dome in the passive containment air-cooling system

Dong Chunhui,Chen Shikang,Chen Ronghua,Tian Wenxi,Qiu Suizheng,Su G.H. 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.8

The Passive containment Air-cooling System (PAS) can effectively remove the decay heat of the modular small nuclear reactor after an accident. The details of natural convection around the dome, which is a key part of PAS, were investigated numerically in the present study. The thermal dynamics around the dome were studied through the temperature, pressure and velocity contours and the streamlines. Additionally, the formation of the buoyant plume at the top of the dome was investigated. The results show that with the increase of Ra, the lift-off point moves toward the bottom of the dome, and the eddy under the buoyant plume grows larger gradually, which enhances the heat transfer. And the heat transfer along the dome surface with different truncation angles was investigated. As the angle increases, the heat transfer coefficient becomes stronger as well. Consequently, a newly developed heat transfer correlation considering the influence of truncation angle for the dome is proposed based on the simulated results. This study could provide a better understanding of natural convection around the dome of PAS and the proposed correlation could also offer more predictive value in the improvement of nuclear safety.