Study on the effect of flow blockage due to rod deformation in QUENCH experiment

Gao Pengcheng,Zhang Bin,Shan Jianqiang 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.8

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat transfer during a reflooding phase and subsequent severe accident. However, most of the system analysis codes simulate the accident process based on the assumed channel blockage ratio, resulting in the fact that the simulation results are not consistent with the actual situation. This paper integrates the developed core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module into the self-developed severe accident analysis code ISAA. At the same time, the existing flow blockage model is improved to make it possible to simulate the change of flow distribution due to fuel rod deformation. Finally, the ISAA-FRTMB is used to simulate the QUENCH-LOCA0 experiment to verify the correctness and effectiveness of the improved flow blockage model, and then the effect of clad ballooning on core heat transfer and subsequent parts of core degradation is analyzed.

Parameter importance ranking for SBLOCA of CPR1000 with moment-independent sensitivity analysis

Xiong, Qingwen,Gou, Junli,Shan, Jianqiang Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.12

The phenomenon identification and ranking table (PIRT) is an important basis in the nuclear power plant (NPP) thermal-hydraulic analysis. This study focuses on the importance ranking of the input parameters when lacking the PIRT, and the target scenario is the small break loss of coolant accident (SBLOCA) in a pressurized water reactor (PWR) CPR1000. A total of 54 input parameters which might have influence on the figure of merit (FOM) were identified, and the sensitivity measure of each input on the FOM was calculated through an optimized moment-independent global sensitivity analysis method. The importance ranking orders of the parameters were transformed into the Savage scores, and the parameters were categorized based on the Savage scores. A parameter importance ranking table for the SBLOCA scenario of the CPR1000 reactor was obtained, and the influences of some important parameters at different break sizes and different accident stages were analyzed.

Study on relocation behavior of debris bed by improved bottom gas-injection experimental method

Teng, Chunming,Zhang, Bin,Shan, Jianqiang Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.1

During the core disruptive accident (CDA) of sodium-cooled fast reactor (SFR), the molten fuel and steel are solidified into debris particles, which form debris bed in the lower plenum. When the boiling occurs inside debris bed, the flow of coolant and vapor makes the debris particles relocated and the bed flattened, which called debris bed relocation. Because the thickness of debris bed has great influence on the cooling ability of fuel debris in low plenum, it's very necessary to evaluate the transient changes of the shape and thickness in relocation behavior for CDA simulation analysis. To simulate relocation behavior, a large number of debris bed relocation experiments were carried out by improved bottom gas-injection experimental method in this paper. The effects of different experimental factors on the relocation process were studied from the experiments. The experimental data were also used to further evaluate a semi-empirical onset model for predicting relocation.

Applicability research of round tube CHF mechanistic model in rod bundle channel

Liu, Wei,Peng, Shinian,Shan, Jianqiang,Jiang, Guangming,Liu, Yu,Deng, Jian,Hu, Ying Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.2

In view of the complex geometric structure of the rod bundle channel and the limitation of the current CHF visualization experiment technology, it is very difficult to obtain the rod bundle CHF mechanism directly through the phenomenon of the rod bundle CHF visualization experiment. In order to obtain the applicable CHF mechanism assumption for rod bundle channel, firstly, five most representative DNB type round tube CHF mechanistic models are obtained with evaluation and screening. Then these original round tube CHF mechanistic models based on inlet conditions are converted to local conditions and coupled with subchannel analysis code ATHAS. Based on 5 × 5 full-length rod bundle CHF experimental data independently developed by Nuclear Power Institute of China (NPIC), the applicability research of each model for CHF prediction performance in rod bundle channel is carried out, and the commonness and difference of each model are comparatively studied. The CHF mechanism assumption of superheated liquid layer depletion that is most likely to be applicable for the rod bundle channel is selected and two directions that need to be improved are given. This study provides a reference for the development of CHF mechanistic model in rod bundle channel.

Study on load tracking characteristics of closed Brayton conversion liquid metal cooled space nuclear power system

Ge Li,Li Huaqi,Shan Jianqiang 한국원자력학회 2024 Nuclear Engineering and Technology Vol.56 No.5

It is vital to output the required electrical power following various task requirements when the space reactor power supply is operating in orbit. The dynamic performance of the closed Brayton cycle thermoelectric conversion system is initially studied and analyzed. Based on this, a load tracking power regulation method is developed for the liquid metal cooled space reactor power system, which takes into account the inlet temperature of the lithium on the hot side of the intermediate heat exchanger, the filling quantity of helium and xenon, and the input amount of the heat pipe radiator module. After comparing several methods, a power regulation method with fast response speed and strong system stability is obtained. Under various changes in power output, the dynamic response characteristics of the ultra-small liquid metal lithium-cooled space reactor concept scheme are analyzed. The transient operation process of 70 % load power shows that core power variation is within 30 % and core coolant temperature can operate at the set safety temperature. The second loop’s helium-xenon working fluid has a 65K temperature change range and a 25 % filling quantity. The lithium at the radiator loop outlet changes by less than ±7 K, and the system’s main key parameters change as expected, indicating safety. The core system uses less power during 30 % load power transient operation. According to the response characteristics of various system parameters, under low power operation conditions, the lithium working fluid temperature of the radiator circuit and the high-temperature heat pipe operation temperature are limiting conditions for low-power operation, and multiple system parameters must be coordinated to ensure that the radiator system does not condense the lithium working fluid and the heat pipe.

Development of mechanistic cladding rupture model for severe accident analysis and application in PHEBUS FPT3 experiment

Pengcheng Gao,Bin Zhang,Jishen Li,Jianqiang Shan 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.1

Cladding ballooning and rupture are the important phenomena at the early stage of a severe accident. Most severe accident analysis codes determine the cladding rupture based on simple parameter models. In this paper, a FRTMB module was developed using the thermal-mechanical model to analyze the fuel mechanical behavior. The purpose is to judge the cladding rupture with the severe accident analysis code. The FRTMB module was integrated into the self-developed severe accident analysis code ISAA to simulate the PHEBUS FPT3 experiment. The predicted rupture time and temperature of the cladding were basically consistent with the measured values, which verified the correctness and effectiveness of the FRTMB module. The results showed that the rising of gas pressure in the fuel rod and high temperature led to cladding ballooning. Consequently, the cladding hoop strain exceeded the strain limit, and the cladding burst. The developed FRTMB module can be applied not only to rod-type fuel, but also to plate-type fuel and other types of reactor fuel rods. Moreover, the FRTMB module can improve the channel blockage model of ISAA code and make contributions to analyzing the effect of clad ballooning on transient and subsequent parts of core degradation.

Design and Implementation of a Novel Horizontal AFM Probe Utilizing a Quartz Tuning Fork

Yifu Chen,Yingzi Li,Guanqiao Shan,Yingxu Zhang,Zhenyu Wang,Mubing Wang,Hua Li,Jianqiang Qian 한국정밀공학회 2018 International Journal of Precision Engineering and Vol.19 No.1

This paper presents a new structure of a novel horizontal atomic force microscope probe utilizing a quartz tuning fork. The horizontal structure exhibits high resistance to environmental noise, and thus the probe can maintain good stability throughout the imaging work. The quartz tuning fork, which is utilized as a force sensor due to its simple mechanical structure and self-actuating and self-sensing characteristics, can significantly simplify the mechanical structure of the probe. The probe is divided into three parts: an approximation device, a force sensor, and a three-dimensional scanner. Each part is carefully designed to guarantee the imaging performance. It is verified that the proposed horizontal AFM probe is stable by conducting finite-element analysis, including modal analysis and noise analysis. Furthermore, the probe is fabricated and the experiments are performed to verify its stability. The proposed horizontal AFM probe combined with the existing control system in the frequency modulation succeeds in imaging within 25 μm × 25 μm and 20 μm × 20 μm ranges stably.

Improvement and validation of aerosol models for natural deposition mechanism in reactor containment

Jishen Li,Bin Zhang,Pengcheng Gao,Fan Miao,Jianqiang Shan Korean Nuclear Society 2023 Nuclear Engineering and Technology Vol.55 No.7

Nuclear safety is the lifeline for the development and application of nuclear energy. In severe accidents of pressurized water reactor (PWR), aerosols, as the main carrier of fission products, are suspended in the containment vessel, posing a potential threat of radioactive contamination caused by leakage into the environment. The gas-phase aerosols suspended in the containment will settle onto the wall or sump water through the natural deposition mechanism, thereby reducing atmospheric radioactivity. Aiming at the low accuracy of the aerosol model in the ISAA code, this paper improves the natural deposition model of aerosol in the containment. The aerosol dynamic shape factor was introduced to correct the natural deposition rate of non-spherical aerosols. Moreover, the gravity, Brownian diffusion, thermophoresis and diffusiophoresis deposition models were improved. In addition, ABCOVE, AHMED and LACE experiments were selected to validate and evaluate the improved ISAA code. According to the calculation results, the improved model can more accurately simulate the peak aerosol mass and respond to the influence of the containment pressure and temperature on the natural deposition rate of aerosols. At the same time, it can significantly improve the calculation accuracy of the residual mass of aerosols in the containment. The performance of improved ISAA can meet the requirements for analyzing the natural deposition behavior of aerosol in containment of advanced PWRs in severe accident. In the future, further optimization will be made to address the problems found in the current aerosol model.

Experimental Investigation on Critical Heat Flux in Bilaterally Heated Annulus with equal heat flux on both sides

Gui Miao,Guo Junliang,Kong Huanjun,Wu Pan,Shan Jianqiang,Peng Yujiao 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.9

A phenomenological study on CHF in a bilaterally heated annulus with equal heat flux on both sides was experimentally performed. The working fluid of the present test was R-134a. Variation characteristics of CHF and transition of CHF occurrence location were investigated under different pressure, mass flux and quality conditions. With the increase of critical thermodynamic quality, it was found that CHF first occurred on the outer surface of the annulus, then simultaneously occurred on both sides, and finally occurred on the inner surface at relatively high critical quality. After the CHF location transitioned to the inner rod, the sharp fall of CHF in the limiting critical quality region was observed. The critical quality corresponding to the CHF location transition decreased with the increase of mass flux and pressure. Besides, CHF in tube, internally heated, externally heated and bilaterally heated annuli were compared under the same hydraulic diameter conditions. The present study is conducive to improving the understanding of complicated CHF mechanism in bilaterally heated annulus, enriching the experimental database, and providing evidence for developing accurate CHF mechanism model for annuli.

Performance evaluation of Accident Tolerant Fuel under station blackout accident in PWR nuclear power plant by improved ISAA code

Zhang Bin,Gao Pengcheng,Xu Tao,Gui Miao,Shan Jianqiang 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.7

The Accident Tolerant Fuel (ATF) is a new concept of fuel, which can not only withstand the consequences of the accident for a longer time, but also maintain or improve the performance under operating conditions. ISAA is a self-developed severe accident analysis code, which uses modular structures to simulate the development processes of severe accidents in nuclear plants. The basic version of ISAA is developed based on UO2eZr fuel. To study the potential safety gain of ATF cladding, an improved version of ISAA, referred to as ISAA-ATF, is introduced to analyze the station blackout accident of PWR using ATF cladding. The results show that ATF cladding enable the core to maintain a longer time compared to zirconium alloy cladding, thereby enhancing the accident mitigation capability. Meanwhile, the generation of hydrogen is significantly reduced and delayed, which proves that ATF can improve the safety characteristics of the nuclear reactor.