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2023 Vol. 44, No. 3

Special Contribution
Study on Hydrogen Induced Delayed Cracking Behavior of Zirconium Alloys Caused by Surface Defects
Zhou Bangxin, Yao Meiyi, Li Qiang
2023, 44(3): 1-7. doi: 10.13832/j.jnpe.2023.03.0001
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Zirconium alloy components used in nuclear reactors maybe occur to fail due to hydrogen induced delayed cracking (HIDC) during service. Whether the microdefects on the surface of the components will cause HIDC is worth studying. In this paper, samples with microcrack defects on the surface of zirconium alloy were prepared by vacuum electron beam welding method, and the corrosion behavior of these microcrack defects during thermal cycling treatment in high-pressure water at 350℃ after corrosion in superheated steam at 400℃ was studied. As the formation of a wedge-like oxide film embedded in the metal at such defects, a tensile stress concentration and stress gradient zone will be formed at the front end of the wedge-like oxide film, resulting in the diffusion and enrichment of hydrogen, and the precipitation of hydrides. In this case, even if the samples are not subjected to external stress and have no residual stresses, these defects will also cause HIDC, leading to defect extension and cracking. Therefore, when designing and manufacturing various zirconium alloy components used in the nuclear reactor core, attention should be paid to how to avoid such defects formed on the surface of such components.
Research Progress of Hearth Monitoring and Fault Diagnosis for Reactor Critical Equipment
Liu Caixue, Luo Neng, He Pan, Liu Jiaxin, Peng Cuiyun
2023, 44(3): 8-20. doi: 10.13832/j.jnpe.2023.03.0008
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The research, development and application of health monitoring and fault diagnosis technology for reactor critical equipment are important guarantee for the safe, reliable and economic operation of nuclear reactors, which also contribute to the improvement of operation and maintenance of nuclear reactors. In this paper, the basic concepts of equipment health monitoring and fault diagnosis are explained. The research status of sensor technology, monitoring technology and health monitoring system of reactor equipment health monitoring are summarized. Furthermore, the research status of fault diagnosis technology based on mechanism model, expert knowledge and data mining are summarized. The purpose of this paper is to provide ideas and methodological references for understanding, deepening, and expanding basic research, technology development, and engineering applications of equipment health monitoring and fault diagnosis.
Reactor Core Physics and Thermohydraulics
Assembly-Homogenized Calculation based on NECP-MCX and Its Application in HPR1000
Qin Shuai, Li Yunzhao, He Qingming, Bai Jiahe, Dong Wenchang, Cao Liangzhi, Wu Hongchun
2023, 44(3): 21-27. doi: 10.13832/j.jnpe.2023.03.0021
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The assembly code based on the Monte Carlo method can handle problems with complicated geometry and avoid the resonance self-shielding calculation in the deterministic assembly code. However, the Monte Carlo assembly code has certain difficulties when generating diffusion coefficient and assembly discontinuous factors. Therefore, the capability for the generation of assembly-homogenized few-group constants is developed based on the continuous-energy Monte Carlo particle transport code NECP-MCX. The cumulative migration area method, which treats the anisotropy of neutron explicitly, is adopted to generate the diffusion coefficient; the fundamental mode approximation is used to consider the effect of neutron leakage on the neutron spectrum; the approach named as the mesh-surface tally is proposed to calculate the corrected assembly discontinuous factors. The mesh-surface tally approach is verified through VERA 2D assembly problems and the NECP-MCX is adopted to simulate the physics start-up test of the home-developed HPR1000 reactor. The results show that the biases of critical boron concentrations, the isothermal temperature coefficients, and the control rod bank worth values satisfy the limited value when compared to the design report. The NECP-MCX has the ability to generate reliable assembly-homogenized few-group constants of which accuracy satisfies the requirement of engineering calculation, laying a solid foundation for the further application of NECP-MCX in next-generation reactors.
Study on Monte Carlo Importance Sampling Method Based on Unstructured Mesh
Shu Hanlin, Cao Liangzhi, He Qingming, Dai Tao, Huang Zhanpeng
2023, 44(3): 28-37. doi: 10.13832/j.jnpe.2023.03.0028
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In order to improve the modeling and calculation accuracy of the traditional Consistent Adjoint Driven Importance Sampling (CADIS) method that relies on structural-mesh finite-difference discrete ordinate (SN) code to determine the importance distribution of particles to further enhance its capability of dealing with complex-geometric deep-penetration problems, a fully automatic unstructured-mesh CADIS method is studied and implemented in this paper, parallel three-dimensional unstructured-mesh neutron-photon-coupled transport code NECP-SUN based on SN method and discontinuous finite element method (DFEM) is developed and embedded into the Monte Carlo code NECP-MCX as adjoint transport solver. The numerical results of the calculations of the HBR-2 benchmark and the fast-neutron fluence rate of the toroidal field coil boxes in the Chinese Fusion Engineering Test Reactor (CFETR) show that the unstructured-mesh CADIS method has stronger adaptability to complex geometry than the traditional CADIS method, and the results obtained are relatively lower in relative statistical error and closer to the measurements; the figure of merit (FOM) is increased by 1~3 orders than that of direct Monte Carlo simulation. Therefore, the unstructured-mesh CADIS method studied in this paper can better handle deep-penetration problems with complex geometry.
Application Research of Neutron Flux Detector in Irradiation Damage of Fast Reactor Structural Materials
Hu Xiao, Chen Xiaoxian, Chen Xiaoliang, Zhang Zhifeng, Zhou Minlan
2023, 44(3): 38-44. doi: 10.13832/j.jnpe.2023.03.0038
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In order to evaluate the radiation damage of fast reactor structural materials, a set of evaluation methods for radiation damage of fast reactor structural materials is proposed in this paper. According to the characteristics of the fast reactor energy spectrum, the irradiation scheme of the neutron flux detector is designed, the characteristics of the detector foil and the section of the reaction channel are analyzed, and seven kinds of fast neutron flux detectors are selected. At the same time, the iterative method is used to develop the spectrum analysis program in Labview platform. Based on the radiation experimental data of the Russian boron carbide module, the spectrum is analyzed, and combined with the calculation of the module cladding dpa of the Lindhard-Robinson model, and compared with the calculated value of SPECTER. The results show that the deviation between the dpa obtained by the experimental method and the calculated value of SPECTER is within 6%, which is in good agreement. In this paper, a complete set of radiation damage evaluation system for fast reactor structural materials is established, which is of great significance to the radiation damage monitoring of structural materials.
Convergence Optimization of 2D MOC/1D SN Method via Generalized Equivalence Theory Based CMFD Acceleration
Kong Boran, Zhu Kaijie, Zhang Han, Hao Chen, Guo Jiong, Li Fu
2023, 44(3): 45-53. doi: 10.13832/j.jnpe.2023.03.0045
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The generalized equivalence theory based CMFD (GET-CMFD) has great convergence behavior and has been successfully applied to the axial NEM based 2D/1D coupling method. However, when applying the high precision SN as the axial solver, the 2D/1D coupling method faces the convergence problem. To solve this problem, in the case of isotropic transverse leakage and anisotropic transverse leakage, the source of the scalar flux of axial nodal discontinuity factor (NDF) and axial modified diffusion factor (MDF) in the GET-CMFD are optimized. At the same time, the application conditions of the transverse leakage splitting method in 2D/1D coupling method are studied systematically. The results show that by adopting the scalar flux from 2D MOC calculation for the axial NDF axial and MDF in the GET-CMFD equation and adopting the latest updated value of leakage term, 2D/1D coupling method can obtain good convergence. The application condition of the transverse leakage splitting method is that the outgoing angular flux is less than 0, which can guarantee convergence and reduce the precision loss. At the same time, using SN calculation twice in a 2D/1D iteration can significantly reduce the number of iterations without increasing the computational load. By improving the GET-CMFD equation, the leakage splitting method and the iterative process, the convergence of 2D/1D coupling method can be improved obviously.
Research on Ray Effect Method Based on Global Factor Correction for First Collision Source
Yang Chao, Li Zhipeng, Yu Tao, Chen Zhenping
2023, 44(3): 54-58. doi: 10.13832/j.jnpe.2023.03.0054
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In the calculation of neutron transport problems with spatially localized source or largely void region, the discrete ordinate method (SN) suffers from ray effects, which cause computational inaccuracies. The first collision source method is often used to alleviate the ray effect to improve the reliability of results. However, this method requires the calculation of the uncollided neutron flux. Generally, which is typically achieved through ray tracing technology based on the grid center method or the grid corner average method, destroying the conservation principle of the number of uncollided neutrons. In this paper, a global factor correction method is proposed to correct the uncollided neutron flux to meet the principle of neutron number conservation. Through the test of Kobayashi shielding calculation benchmark problem, the calculation results show that the maximum error is reduced from 6.15% to 3.71%, indicating that the method can effectively improve the accuracy of calculation results and provide data support for shielding optimization design.
Multi-scale Coupling Transient Characteristics of Xi 'an Pulsed Reactor Based on RELAP5 and CTF
Zhao Wei, Chen Lixin, Liu Guodong, Zhang Liang, Yang Ning, Zhang Xinyi, Tian Xiaoyan, Sun Peiwei, Yuan Jianxin
2023, 44(3): 59-64. doi: 10.13832/j.jnpe.2023.03.0059
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Based on the system code RELAP5 and the sub-channel code CTF, a multi-scale coupling domain decomposition approach for pool pulsed reactor with natural circulation and a grid mapping method that meets the precise conversion between different scale parameters are proposed in this paper. The correctness and feasibility of the coupling method are verified by the analysis of the steady state and transient conditions of the pulsed reactor. The results show that the proposed multi-scale coupling method can be applied to the coupling analysis of the pulsed reactor. The results of coupling calculation are better than those of single system code in accuracy. The accuracy of the fule core temperature is in creased by approximately 62.50% in the 2 MW steady-statecondition and 76,71% under the pulsed transient condition. Meanwhile, the high-precision and high-resolution thermal-hydraulic distribution and transient characteristics of the core can be obtained by coupling calculation.
Research on Prediction of Subcooled Flow Boiling CHF for Spiral Flow Based on Machine Learning
Yan Jianguo, Zheng Shumin, Guo Pengcheng, Zhao Li, Wang Shuai, Liu Kun, Zhu Xutao
2023, 44(3): 65-73. doi: 10.13832/j.jnpe.2023.03.0065
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Subcooled boiling is widely used in cooling applications with high heat fluxes represented by International Thermonuclear Experimental Reactor (ITER). In this paper, predictions on the critical heat flux (CHF) of subcooled water boiling under high heat flux and swirl flow conditions are focused and a database of subcooled boiling CHF is established. Machine learning methods are applied, in which four typical machine learning models are adopted, namely Back Propagation (BP) neural network, Genetic Algorithm (GA)-BP neural network, Radial Basis Function (RBF) neural network and Extreme Learning Machine (ELM). The results indicate that machine learning models can effectively predict subcooled boiling CHF with swirling flow, and the prediction performances are obviously promoted, comparing with traditional empirical correlations. Among typical machine learning models, the ELM model possesses the best performance, with the MAE and RMSE are equal to 2.79% and 4.22%, respectively. The results provide a new path to making accurate predictions on the CHF for subcooled boiling under high heat flux and swirling flow conditions.
Study on Effect of Temperature on the Narrow Gap Ressistance Coefficient Test
Meng Yang, Fan Ruichen, Sui Xi, Li Yong, Zheng Jiantao, Zhang Jiaqi, Wang Jie, Yu Ting
2023, 44(3): 74-78. doi: 10.13832/j.jnpe.2023.03.0074
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The coolant leakage and bypass flow channel in PWR is mostly a narrow gap. The size of narrow gap is sensitive to be easily affected by system pressure, differential pressure, temperature, vibration and other factors. Small size changes will cause significant changes in resistance coefficient, resulting in the measured Reynolds number and resistance coefficient curve contrary to the basic principles of fluid dynamics. In this paper, the study on effect of temperature change on the flow resistance coefficient in narrow gap is conducted under the conditions of constant temperature and temperature rise. During the temperature rise test, the system pressure, the differential pressure and flow of the test body stays unchanged, and gradually increase the temperature of the fluid, then obtain the resistance coefficient of the test body under different temperature. The test results show that when the temperature of the test fluid rises from 23°C to 52°C, the narrow gap of the test body is more closely fitted due to thermal expansion and cold contraction, and the flow resistance coefficient increases by 8%.
Analysis on Thermal Hydraulic Characteristic of Helical Coiled Tube Steam Generator of Liquid Metal Fast Reactor Based on Drift-Flux Model
Liu Jialun, Li Huixiong, Zhang Shunzhe, Ning Liang, Tang Linghong
2023, 44(3): 79-89. doi: 10.13832/j.jnpe.2023.03.0079
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In the helical coiled tube steam generator of liquid metal fast reactor, the temperature difference between inlet and outlet of the primary side increases significantly, and the outlet steam superheat degree of the secondary side also increases significantly. This is a common problem and brings new challenges to the safe operation of steam generator. Based on the discrete grid method, a computational model suitable for the thermal-hydraulic characteristics of the helical coiled tube steam generator of the liquid metal fast reactor was established. The model meshed both the primary side circuit and the secondary side circuit. The drift-flux method was adopted to calculate the flux and heat transfer process of two-phase steam-water flow along the helical coiled tube, and the correlations of liquid metal physical properties and liquid metal heat transfer were selected for the primary side calculation. Meanwhile, the inner node method was adopted to divide the tube wall into a series of grids, and the heat conduction equations were built to accurately simulate the convective heat transfer between the fluid on both sides and the tube wall, as well as the wall metal heat conduction process. The present model was then verified based on the experimental data. Finally, taking the lead-bismuth fast reactor as an example, the thermal-hydraulic characteristics of the helical coiled tube steam generator were analyzed under different inlet conditions. It was found that the wall heat flux distribution between the primary and secondary sides is extremely uneven along the helical coiled tube steam generator, and the peak wall heat flux is extremely high. The maximum value of the wall heat flux reaches 1,361 kW/m2, and the difference between the maximum value and the minimum value is tens to hundreds of times in the example of this paper. With the increase of lead-bismuth temperature and velocity at the inlet of primary side, the lengths of the subcooled water zone and the two-phase zone on the secondary side are significantly shortened, and the length of the superheated steam zone is significantly increased. Meanwhile, the peak wall heat flux moves towards the inlet of the helical coiled tube, and the total pressure drop of the working medium on the secondary side also increases significantly.
Experiment Research on the Mechanism of Fuel Migration Behavior under Severe Accident of Lead-based Reactor
Cai Qinghang, Chen Ronghua, Xiao Xinkun, Fu Hao, Tian Wenxi, Qiu Suizheng, Su Guanghui
2023, 44(3): 90-95. doi: 10.13832/j.jnpe.2023.03.0090
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In the severe accident of lead-based reactor, the fuel particles may lead to core blockage during the migration process in the reactor, and there is a risk of re-criticality. In this study, in order to obtain the flow and migration characteristics of metal particles entrained in liquid lead-bismuth eutectic (LBE), the visualization experimental section of quartz glass was designed, and the visual experimental device called Eirene for the flow and solidification behavior of LBE entrained particles was built. And the flow solidification experiment of LBE in a circular channel and that of LBE entrained stainless steel particles were carried out. The image data of flow and solidification characteristic of LBE, the temperature data of inlet, outlet and tube wall were obtained. The results show that the existence of metal particles seriously hinder the flow of LBE in the tube. It is easy to form solidification blockage in the dense area of the particle bed. A large number of particles are anchored in the middle of the experimental section and gradually form blockage from the outside to the inside. The experimental results can provide support for the verification of the lead-based reactor severe accident analysis models.
Nuclear Fuel and Reactor Structural Materials
Study on Oxidation Corrosion Characteristics of Horizontal Lead-Bismuth Reactor Core
Lu Dingsheng, Wang Chen, Wang Chenglong, Yue Nina, Yang Ping, Tian Wenxi, Su Guanghui, Qiu Suizheng
2023, 44(3): 96-103. doi: 10.13832/j.jnpe.2023.03.0096
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In order to study the oxidation corrosion characteristics of horizontal lead-bismuth reactor core, a liquid lead-bismuth oxidation corrosion model was established in this study. Based on computational fluid dynamics method, the self-defined source term method of transport equation was used to realize coupling calculation. The results show that the thickest oxidation layer on the fuel rod surface of the reactor core is located at the outlet position under the reference condition, and the oxidation layer on the fuel rod surface at the center is significantly higher than that near the fuel assembly box. After 10,000 hours, the surface of the fuel rod at the center still maintained a double oxide layer structure, and the average total thickness of the double oxide layer was 1.32 μm. This study provides a numerical simulation method for the oxidation corrosion characteristics of lead-bismuth reactor cores, which can be used for the prediction of oxidation corrosion of lead-bismuth reactor cores.
Structure Design and Performance Analysis of 90Sr Radioisotope Thermoelectric Generator
Du Guanghan, Li Yupeng, Li Gen, Guo Rui, Liu Guixiu, Wang Jinshi
2023, 44(3): 104-111. doi: 10.13832/j.jnpe.2023.03.0104
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In this paper, a physical model of 90Sr radioisotope thermoelectric generator was designed, including the heat source, thermoelectric conversion module and heat dissipation module. On this basis, the finite element analysis of generator was carried out by using the COMSOL, and the temperature distribution, voltage and power output characteristics of generator were obtained. Meanwhile, the output characteristics of generator under the attenuation of the heat source power were studied, and finally the sensitivity analysis of the height and cross-sectional area of the thermoelectric material was carried out. The research results show that the maximum output power of generator is 63.6 W, the thermoelectric conversion efficiency is 7.6%, and the maximum output power of generator is reduced from 63.6 W to 24.4 W within 15 years. Sensitivity analysis found that the internal resistance and maximum output power of generator increase with the increase of the height of the thermoelectric material, and decrease with the increase of the cross-sectional area. The research results are of great significance to the application of radioisotope thermoelectric generator in harsh environments such as space and deep sea.
Identification of Flow Regime of Boiling Flow in a Vertical Annulus with Unsupervised Machine Learning
Zhu Longxiang, Zhang Luteng, Sun Wan, Ma Zaiyong, Pan Liangming
2023, 44(3): 112-120. doi: 10.13832/j.jnpe.2023.03.0112
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Accurate flow regime identification of the boiling flow is of great significance for using closure correlations in thermal-hydraulic system codes. The flow regime identification is achieved by the unsupervised machine learning (ML) approach, and the current work integrates the data-driven ML model with the two-phase-flow domain knowledge. Two requirements are established to determine whether the feed-in data type is appropriate or not: ①the information captured by the unsupervised ML should be regime-relevant features for distinguishing flow regimes;②the cluster criterion for a flow regime should cover all the possible feed-in features of the regime. Feed-in data types generated by the conductivity probe are examined; among them, only the bubble chord length Cumulative Distribution Function (CDF) fulfills the two requirements. With the feed-in data of chord length CDF, the two-dimensional local flow regimes are identified and analyzed for a boiling dataset conducted in an internally heated vertical annulus. The results show that the higher-rank regimes appear at the channel’s center leaning toward the inner-heated wall. The global flow regime map is obtained with the local regimes, and a new flow regime transition criterion between the bubbly and slug flow is developed as the void fraction equals 0.14.
Study on the Microstructure and Mechanical Properties of TRISO Microsphere
He Zongbei, Chen Jiangshan, Zeng Qiang, Pan Xiaoqiang, Yao Lifu, Qiu Shaoyu
2023, 44(3): 121-126. doi: 10.13832/j.jnpe.2023.03.0121
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The TRISO microspheres were prepared by fluidized-bed chemical vapor deposition (FBCVD) process with ZrO2 microspheres as simulated cores. The microstructures of the coating layers of the microspheres were inspected by SEM and TEM, and the elastic modulus as well as Vickers hardness of the coating layers were tested by micro indentation. Furthermore, the fracture strength of the SiC layer was tested by the crushing test. The results indicate that: ①the Buffer layer of the TRISO microspheres is composed of a large number of spherical particles, and there are many large pores in the coating layer; ②the microstructures of the IPyC and OPyC layers are similar, which are composed of a large number of spherical carbon particles coated with lamellar structure, with relatively high density; ③ the SiC layer exists in a typical β-SiC crystal state, and the atomic layer spacing is 0.26 nm; ④among all the coating layers, the Buffer layer has the lowest elastic modulus and hardness, which are about 13.29 GPa and 1.78 GPa respectively; however, the elastic modulus and hardness of inner PyC layer are about 25.80 GPa and 3.18 GPa respectively, and the elastic modulus and hardness of outer layer PyC are about 28.15 GPa and 3.66 GPa respectively; ⑤ for SiC layer, it has the highest elastic modulus and hardness, which are about 141.4 GPa and 21.3 GPa respectively, and its average fracture strength is 2581 GPa. The different structures of TRISO microspheres coatings come from different deposition process, and the thermal decomposition reaction of precursor gas is the key to controlling the coating structure. This paper studies the structure and mechanical characteristics of the TRISO microspheres, and obtains corresponding performance data, which has important guiding significance for the design and application of TRISO microspheres and the prediction of in-reactor radiation.
Research on Hoop Mechanical Property of Zr-Sn-Nb Alloy Cladding Tubes in Vacuum
Yan Meng, Li Shunping, Wang Pengfei, Hong Xiaofeng, Liang Bo, Dai Xun, Yin Qiwei
2023, 44(3): 127-131. doi: 10.13832/j.jnpe.2023.03.0127
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At present, there is no reference standard for hoop tensile testing of small-diameter tubes and there is also no uniform research method for hoop tensile testing of zirconium alloy cladding tubes in the worldwide research field. As to home-made Zr-Sn-Nb alloy cladding tubes, there is no data in hoop mechanical property. In this paper, the hoop tensile mechanical properties of Zr-Sn-Nb alloy cladding tubes in vacuum were obtained by using the hoop specimen with a gauge section. The results show that hoop tensile strengths and elongations dropped with temperature rise from room temperature to 600℃, but there is an abnormal phenomenon with flat stage in strength at 300-400℃ due to the influence of the dynamic strain aging, the hoop tensile strengths are lower than the axial tensile strengths at the same temperature in vacuum, and the gap widened as the temperature increased; the hoop tensile curve at 550℃ and 600℃ jitters wavelike due to the influence of dynamic recrystallization; the fracture characteristics vary significantly at different temperatures, indicating that the hoop tensile fracture mode depended on the temperature heavily.
Structural Mechanics and Safety Control
The Study on Seismic Analysis and Test Method of Nuclear Grade Fan
Zhang Xiaowei, Zhou Bin, Niu Haoxuan, Yu Yang
2023, 44(3): 132-137. doi: 10.13832/j.jnpe.2023.03.0132
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In this paper, the seismic analysis and seismic test of a nuclear vortex process fan are carried out. Based on the test, the seismic analysis method of this kind of fan is obtained. Firstly, the finite element model of the fan is established, and the main modal frequencies of the fan are obtained through modal analysis, which are compared with the dynamic characteristic detection test results; Then, the strain of the fan is analyzed and calculated by the response spectrum method, and compared with the strain results measured in the seismic test. The results show that the seismic analysis method used in this paper is reasonable and can obtain accurate seismic response results, so it can be considered as an alternative to the seismic test of this kind of fan.
Analysis of Magnetic Saturation Characteristics of Control Rod Drive Mechanism
Yang Yun, Yu Tianda, Chen Yangming, Li Qingzhao, Liu Yanting, Xu Qiwei, Luo Lingyan
2023, 44(3): 138-143. doi: 10.13832/j.jnpe.2023.03.0138
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In this paper, the impact of magnetic saturation on the electromagnetic parameters and magnetic circuit parameters of the control rod drive mechanism (CRDM) is studied by combining theoretical and finite element analysis (FEA). The method of obtaining magnetic circuit parameters is derived by establishing an equivalent magnetic circuit (EMC) model of the CRDM. The change law of the electromagnetic and magnetic circuit parameters of the CRDM with the current and length of the main air gap is obtained based on FEA. The results show that magnetic saturation significantly affects the main magnetic circuit's residual reluctance and inductance. The flux leakage coefficient of the CRDM has little relationship with the magnetic saturation but is greatly affected by the length of the main air gap. The reluctance of the main air gap should consider the air gap flux edge effect. This study provides a theoretical basis for the EMC model optimization of CRDM.
Research on Heat Pipe Reactor Startup Process based on Autonomous Operation
Liu Yu, Huang Mengqi, Peng Changhong, Du Zhengyu
2023, 44(3): 144-151. doi: 10.13832/j.jnpe.2023.03.0144
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The application of the heat pipe reactor (HPR) urgently requires unmanned autonomous operation technology. Applying autonomous operation technology to HPR can realize state sensing, trend prediction and strategy optimization, which can effectively avoid human errors, improve the technical performance of HPR and expand the application fields of nuclear power. In this paper, the MegaPower reactor was used as the research object and the HPRTRAN program was used as the analysis tool to carry out an autonomous operation research based on the HPR start-up process, which is an important component of the HPR operation process, and then an autonomous operation framework consisting of monitoring and diagnosis layer, prediction layer and decision layer was established for the HPR start-up. The research results show that the prediction results of the autonomous operation system are highly accurate and the decision-making scheme is scientific and feasible. The research results can lay a foundation for the subsequent full realization of the unattended autonomous operation of the HPR.
Quantitative Research on Parameter Uncertainty of Helical Coil Once-Through Tube Steam Generator on the Small Pressurized Water Reactor
Liang Lehua, Wang Xuejian, Zeng Wenjie, Li Chuhao
2023, 44(3): 152-159. doi: 10.13832/j.jnpe.2023.03.0152
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To study the influence of parameter uncertainties of helical coil once-through tube steam generator (HCOTSG) on the operation of small pressurized water reactor, based on the establishment of the core power control system, the feed water control system of steam generator and the pressure and water level control system of pressurizer a quantitative platform for parameter uncertainties of HCOTSG on the small pressurized water reactor was developed. Taking the uncertainty quantification of inner and outer diameters of helical pipes as an example, the application research of the platform is carried out. The results show that the developed platform can well quantify the uncertainty of the primary system of small pressurized water reactor.
Improvement and Analysis of Design of Secondary-side Passive Residual Heat Removal System for PWR
Xian Lin, Li Feng, Yu Na, Wu Qing, Qiu Zhifang, Deng Jian, Lu Yili, Li Haiying
2023, 44(3): 160-164. doi: 10.13832/j.jnpe.2023.03.0160
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In applying the design of secondary-side passive residual heat removal system (PRS) of Generation 3 PWR to Generation 2+PWR, there are limitations due to the long distance between the cooling water tank of the PRS with the shell of the PWR, which complicates the arrangement of the relatively long steam pipe and condensation pipe of the PRS. This paper presents an improved design for PRS. The ARSAC code was used to simulate and calculate the transient process of the system under different operating conditions, namely, the initial isolation valve of the steam pipeline was closed and the pipeline was filled with nitrogen, steam, and water, respectively. The results were compared with those obtained under the initial condition of an open valve. The comparison results show that all designs can meet the requirement of residual heat removal, but have advantages and disadvantages in engineering feasibility and system operational stability. Therefore, the improvement design with the steam pipe are full of nitrogen is most recommended to apply in engineering.
Development and Application of Control Rod Drive Mechanism Expansion Reactivity Feedback Model
Zhang Xisi, Yang Peng, Xue Fangyuan, Huo Xingkai, Liu Yizhe
2023, 44(3): 165-168. doi: 10.13832/j.jnpe.2023.03.0165
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In the power test phase of China Experimental Fast Reactor, the power reactivity measurement test was carried out. The results showed that there was a large deviation between the measured values and the theoretical calculation values, and the calculation method of power reactivity needed to be improved. This research analyzed the deviation between the theoretical calculation values and the measured values of China Experimental Fast Reactor power reactivity, and the cause of the deviation was found out. Furthermore, this research developed a control rod drive mechanism expansion reactivity feedback model to correct the theoretical calculation results. The results show that the corrected theoretical calculation values are in good agreement with the measured values, with a relative error of 6.5%. Therefore, the control rod drive mechanism expansion reactivity feedback model developed in this research can be used for reactivity calculation in sodium cooled fast reactors.
Circuit Equipment and Operation Maintenance
Research on Assembly Process of HTR-PM Steam Generator Insulation Materials
Wang Yin, Pu Chenghao, Wang Qiang
2023, 44(3): 169-172. doi: 10.13832/j.jnpe.2023.03.0169
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In order to solve the assembly problem of insulation materials for steam generator of high temperature gas cooled reactor demonstration project (HTR-PM) and ensure that the insulation materials meets meet the design requirements of filling density and pressure, the assembly process test was carried out at different positions of the steam generator. The filling density of insulation material under different pressure was obtained by measuring the relationship between compression and rebound through test. The results show that blanket insulation material is used on the hot helium side and conical block and flocculent insulation material on the cold helium side, which can meet the density and pressure requirements of the design. Therefore, the assembly problem of steam generator insulation materials has been successfully solved through this research, which provides technical guarantee for the subsequent construction of high temperature gas-cooled reactor projects.
Study on Startup Characteristics and Operation Limits of Medium-Temperature Dowtherm Heat Pipe
Xue Zhihu, Lu Qin, Qu Wei, Liu Chao, Yu Jijun
2023, 44(3): 173-179. doi: 10.13832/j.jnpe.2023.03.0173
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Based on the needs of passive heat removal system using medium-temperature heat pipes (MTHPs) and heat pipe heat exchangers (HPHEs) in the high temperature circuit of large nuclear power plant and the novel nuclear reactor, and promoting their applications in industrial heat transfer and energy conservation at medium-grade temperature zone, a study of start-up characteristics and operation limit features of a MTHP (using Dowtherm-A as working medium) was conducted through the experimental method of testing the temperature response. The startup test results show that the startup temperature of the Dowtherm-A heat pipe (DHP) is about 200℃ and the DHP can start up completely before 250℃. The limit test results demonstrate that the operation limit temperature of the DHP is 497℃ and it will work stably over 400℃ if its condenser is well cooled.
Development of Online Sipping Detection Device Based on β-γ Coincidence Method
Zeng Yong, Gu Mingfei
2023, 44(3): 180-184. doi: 10.13832/j.jnpe.2023.03.0180
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Abstract:
The fuel cladding tubes damage often occurs during the operation of the reactor. When a fuel assembly is damaged, the power plant usually performs a sipping test on each fuel assembly using an online sipping device during the re-fueling process, and determines whether the tested assembly is damaged. However, due to the influence of the strong radiation in the island, misjudgment and missed judgment may occur in the detection results by detecting single β or γ particle. In this paper, an online-sipping detection device based on β-γ coincidence measurement technology is designed, which contains β and γ particle detectors, and a β-γ coincidence detection channel. It combines existing β and γ damage judgment technologies and adds the results based on β-γ coincidence judgments. The results of the applications show that the device background counting rate can reach 0.04 s–1 and the Minimum Detectable Activity (MDA) of 133Xe can reach 23.7 Bq under the environmental radiation in the nuclear island. The detection sensitivity is better than that of a single β or γ particle detection mode. The application of the three judgment modes significantly enhances the reliability of damage determination results. This device can replace the existing online sipping detection device of single particle detection mode and be applied to the nuclear power site to carry out relevant detection work.
Study on T2 Test Time Optimization Method of DAS
Zhang Ruixia, Wang Dong, Li Gang, Chen Yongqun, Wang Jikun, Han Bin
2023, 44(3): 185-188. doi: 10.13832/j.jnpe.2023.03.0185
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Abstract:
In the logic function verification test (T2 test) of the Diversity Actuation System (DAS), the verification of the timer in the logical link is often carried out by waiting, making the test last for more than 2 hours, causing human error in the waiting process. In order to reduce T2 test time and human failure, in this study, an optimization method of T2 test time for DAS based on simple hardware technology was proposed. When verifying the logic function, the timers were bypassed, and then the timers were accelerated by adjusting the reference clock of them. The result shows that the total time of T2 test for DAS can be controlled within 30 minutes. Therefore, compared with the traditional test method, the test time is greatly shortened, and the human failure during the test can be effectively avoided, so as to meet the latest requirements of T2 test time in nuclear power plants.
Research and Application of Significance Determination for Safety-Related Feedback Information Based on Analytic Hierarchy Process in Nuclear Power Plants
Zou Xiang, Zhang Hao, Lei Lei, Xu Youlong, Zheng Lixin, Jiao Feng
2023, 44(3): 189-195. doi: 10.13832/j.jnpe.2023.03.0189
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Abstract:
The operating experience feedback can significantly improve the safety level and operation level of nuclear power plants. China’s nuclear industry regulatory agencies and various operating units have established different levels of experience feedback systems that are operating well, collecting large amounts of experience feedback information. However, it is hard to evaluate the importance of these information and chose more valuable information for in-depth feedback because of the shortage of existing methods. The paper developed and practiced an experience feedback importance evaluation method based on the hierarchical analysis process (AHP), which identified five experience feedback main elements and seven sub-elements to conduct a comprehensive and objective evaluation of information. Ultimately, quantitative feedback value scores are provided to identify important experience feedback information and to improve the nuclear power plant experience feedback work of China.
Experimental Investigation on Measuring the Leakage Rate of Nuclear Power Plant Containment by Pressure Drop Method
Li Jianfa, Liu Feng, Hua Yongzhen, Sun Zhongning, Meng Zhaoming
2023, 44(3): 196-201. doi: 10.13832/j.jnpe.2023.03.0196
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Abstract:
In order to further explore the basic law that the environmental conditions in the nuclear power plant containment affect the leakage rate measurement, a data analysis program for measuring the containment leakage rate by the pressure drop method was developed in this investigation. The measurement results were validated by leakage rate measurement tests carried out on a large-scale containment simulant. Based on the test data, the effects of the temperature and relative humidity on the leakage rate measurements under quasi-steady-state and non-steady-state conditions were studied. The results show that in the heating environment, the different heating rates of each temperature sensor due to the difference in the thermal environment can significantly affect the measurement of the leakage rate, and the relative deviation of the measured value compared with the temperature before heating can exceed 20%. Moreover, in the unsteady environment of local temperature and humidity changes in the vessel, the inherent delay effect of the pressure and temperature sensors can also cause abnormal leakage rate measurements. However, when the temperature and humidity in the casing are kept constant or the temperature and humidity change rates of each sensor are approximately the same, the measured value of the leak rate is relatively stable. The conclusions provide a reference for further understanding the leakage rate measurement mechanism and method optimization.
Research on Calculation of Coolant Radiolysis Products in Operating Conditions during the Shutdown of PWRs
Li Fuhai, Lin Yunliang, Lin Zijian, Lin Genxian, Guo Zifang, Fang Jun, Lin Mingzhang
2023, 44(3): 202-209. doi: 10.13832/j.jnpe.2023.03.0202
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Abstract:
The control of radioactive source of the activated corrosion products during the shutdown of PWRs is one of the most effective means to reduce the CRE. The effect of coolant radiolysis should be taken into account. In this paper, a coolant radiolysis kinetic model was developed based on water radiolysis reaction kinetics. The main radiolysis products such as H2O2, O2 and H2 under different conditions were obtained through the kinetic calculations. The reaction mechanism was also discussed. The results showed that: ① during the shutdown, as the temperature of coolant decreases from 310℃ to 60℃, and the concentration of dissolved H2 decreased to 0, the coolant changes from reductive to oxidative. The radiolysis generation of H2O2 and O2 is significantly promoted;② the generation of radiolysis products is affected by the chemical parameters, such as concentrations of B/Li, dissolved H2, dissolved O2 and H2O2; ③ the generation of H2O2 and O2 are mutually promoted by each other. The presence of H2 can inhibit the generation of H2O2 and O2 to some extent. The presence of high concentration of H2O2 will protect H2 from the consumption of ·OH. The results in this paper may provide important references and guidances for the selection of operating parameters under different conditions during the shutdown of PWRs and the control of activated corrosion products.
Anomaly Detection of Core Self-Powered Neutron Detector Based on Twin Model
Chen Jing, Lu Yanzhen, Jiang Hao, Lin Weiqing, Xu Yong
2023, 44(3): 210-216. doi: 10.13832/j.jnpe.2023.03.0210
Abstract(260) HTML (107) PDF(43)
Abstract:
Self-powered neutron detector (SPND) is an important nuclear sensing device in the core, whose health status affects the safe operation of the reactor directly. Considering the measurement correlation between SPNDs at different positions in the reactor, a twin model-based anomaly detection method for SPND signals in the core was proposed in this paper. The characteristics of measurement signals of neighboring SPNDs were extracted by the Random Forest Regression (RFR) algorithm based on the historical operation data of SPND, and a twin model was built for SPNDs which outputs the same as its physical entity. Twin model and entity sensing coexisted. The residual error between the actual observation value of SPND and the twin model estimation value was calculated to serve as the anomaly detection criterion, which realized the identification and location of single-point and multi-point SPND anomalies. The experiments show that the prediction error of the twin model proposed in this paper attains the order of 1×10−10, which has a very high output consistency. The identification accuracy of anomaly detection can reach over 99% under various abnormal states of SPND signals, and the single-point and multi-point abnormal SPND can be accurately identified, which has a high reference value for improving the reliability and safety of the state monitoring of neutron flux measurement system in the core.
Study on the Steady γ-Radiolysis of Ammonia Solution
Guo Zifang, Yang Yu, Lin Mingzhang, Cao Qi, Tang Jia, Lin Yunliang, Lin Zijian
2023, 44(3): 217-222. doi: 10.13832/j.jnpe.2023.03.0217
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Abstract:
Ammonia is applied in the coolant to eliminate the O2 and H2O2 in the PWR, thus mitigating the corrosion of structural materials. The present work studied the γ-radiolysis of ammonia solution under different conditions including ammonia concentration, absorbed dose, absorbed dose rate, gas-liquid volume ratio, and different saturated gases. The results show that ammonia significantly inhibits H2O2 because of the consumption of H2O2 and suppression of its precursors. The concentration of NO2 rises with the ammonia concentration . Ammonia is continuously consumed as irradiation progresses, while the concentration of H2O2 increases significantly with the absorbed dose. The NO2 reaches a peak maximum (> 100 μmol/L) when the absorbed dose is 8 kGy. However, NO2 has a drop when the absorbed dose grows further, due to the fact O2 oxidizes NO2 into NO3. The concentrations of H2O2 and NO2 are not obviously affected within the absorbed dose rate range (2.78~25 Gy/min). The presence of O2 is critical to the formation of NO2, though excessive O2 in the system could lead to the oxidization of NO2 by increasing the oxidizing species. Besides, the oxygen dissolved in aqueous ammonia promotes the production of H2O2. This work is expected to provide a helpful reference for the optimization of the ammonia-containing coolant system.
Column of Science and Technology on Reactor System Design Technology Laboratory
Numerical Simulation of Flow Heat Transfer in Rectangular Channel under Local Deformation Condition
Chen Mingrui, Wei Zonglan, Chen Chong, Deng Jian, Zhu Li, Peng Shinian
2023, 44(3): 223-230. doi: 10.13832/j.jnpe.2023.03.0223
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Abstract:
The complex working environment of the fuel elements in the reactor may cause changes in the performance of fuel elements and deviations from their initial geometric state, which can affect flow and heat transfer characteristics and threaten the safety of the reactor core. In this paper, the ANSYS Workbench numerical simulation platform was used to establish a model containing four coolant channels, and the steady-state numerical simulation was carried out considering different bending conditions in the solid domain. The results show that the coolant flow is redistributed among the four channels under different bending conditions, thus affecting the temperature in the channel with a small flow section area increases significantly, and the highest temperature point in the solid domain shifts from the central region to the channel with a reduced flow area.
The Implementation and Efficiency Analysis of Parallel Mesh Mapping based on the Multi-physics Coupling Framework
Tang Qifen, Wang Yuan, Pan Junjie, Qiang Shenglong, Fan Jiakun, Cui Xiantao
2023, 44(3): 231-236. doi: 10.13832/j.jnpe.2023.03.0231
Abstract(340) HTML (44) PDF(71)
Abstract:
Refined physical-thermal coupling calculations of reactors can simulate the core behavior more accurately. However, existing analysis programs adopt different discrete formats and mesh divisions when calculating different physical fields, resulting in a complex mesh mapping relationship for the transfer of discrete variables between physical fields. Especially for the refined whole-core modeling, the large-scale mesh mapping will affect the accuracy and efficiency of the coupled system solution. In this paper, based on the self-developed multi-physics coupling framework MORE, the thermal-hydraulic sub-channel software CORTH integrated in MORE, and the Monka program RMC, the whole core refined mesh of the physical-thermal coupling calculation was realized by the method of area decomposition parallel mesh mapping. The mapping time between the million-level structured mesh and unstructured mesh can be reduced to 8 s on 20 cpus, and maximum parallel mapping efficiency reaches 77.96% on 10 cpus, which improves the coupling calculation efficiency.
Characterization Analysis of Core Outlet Temperature Measurement in Nuclear Power Plants
Xin Sufang, Wu Dan, Shen Yaou, Ren Chunming, Chen Shilong
2023, 44(3): 237-242. doi: 10.13832/j.jnpe.2023.03.0237
Abstract(180) HTML (71) PDF(31)
Abstract:
The core outlet temperature measurement is of great significance for mastering the reactor operation status. This paper analyzed the characterization of the core outlet temperature measurement by computational fluid dynamics (CFD) method. By simulating the structure of the fuel assembly and instrument tube, the coolant flow and temperature distributions in the instrument tube were obtained. The relationship between core outlet temperature characterization and fuel assembly power was obtained through quantitative analysis of core outlet temperature measurement results under 9 typical power distributions. The calculation results show that the average temperature of the measurement point is basically linearly related to the power of the fuel rod, and the temperature value of the measurement point increases with the increase of the power of the fuel rod, while the characterization of the temperature measurement becomes worse with the increase of the power of the fuel rod. The research results provide a certain basis for the calibration of outlet temperature measurement.
Multi-fidelity Mechanical Model Matching and Updating of Primary Loop Supporting Structure
Xiong Furui, Zhang Wenzheng, Yuan Zhihao
2023, 44(3): 243-248. doi: 10.13832/j.jnpe.2023.03.0243
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Abstract:
The design of reactor primary loop support should meet the vibration reduction and anti-shock requirements. Accordingly, mechanical models with different fidelity should be established. This paper first adopted correlation analysis to quantify the compatibility of static and dynamical characters of multi-fidelity mechanical models (including high-fidelity and low-fidelity models) of the reactor and primary loop supporting structure. Then, Bayesian parameter estimation technique was applied to update the low-fidelity model (simplified model) to increase its compatibility to high-fidelity model (detailed model). The model matching and updating strategy proposed by this paper lays the foundation for future multi-objective optimization of support structure mechanical properties.