核动力工程

Evaluation of Component Cooling System based on Optimization Algorithm in HPR1000

Yu Pei, Hou Ting, Zhao Weiguang

2022, 43(S2): 1-6. doi: 10.13832/j.jnpe.2022.S2.0001

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Abstract:
Aiming at the problems of large design margin and low winter supply water temperature of cold chain systems such as equipment cooling water in power plants, an innovative design method is provided, which can be used to determine the system configuration and operation scheme in the overall design stage. Firstly, based on the basic principle of system heat balance, a thermal evaluation model is established, and an economic evaluation model is established based on the guiding ideology of improving economy as much as possible under safety and operating conditions. Then, based on the basic principle of optimization design, the multi-objective optimization algorithm and analysis program are developed. Finally, the above program is used to deal with the multi-objective optimization problems in the system design, and finally the system scheme evaluation is realized. The optimization evaluation results of the cooling water system of HPR1000 equipment are given, which significantly improves the economy of the power plant. It breaks through the traditional single-line design process which relies too much on design experience and is subject to insufficient quantitative analysis. In the adjustment of multi-discipline and multi-variable design scheme and parameter optimization, software decision-making is used to assist manual decision-making to shorten the design time and improve the accuracy.

Study on the Application of Spectral Shift Absorber in Special Criticality Safety

Gao Jian, Su Yi, Zhao Runzhe, Guo Jian

2022, 43(S2): 7-12. doi: 10.13832/j.jnpe.2022.S2.0007

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Abstract:
Special criticality safety is a key issue in the design of space reactors, especially fast reactors. Once a fast neutron space reactor falls into water, dry sand or wet sand due to launch drop accident, the neutron moderating ability and reactivity will increase, which may lead to a criticality safety accident of the reactor. During the design, it is necessary to ensure that the reactor remains subcritical under all drop accidents. Spectral shift absorber (SSA) is a kind of material with much larger neutron absorption cross section in resonance energy region than in fast neutron energy region. Its application in fast neutron space reactor can significantly improve the criticality safety performance of fast neutron space reactor. Taking the fast neutron space reactor cooled by the liquid metal circuit as the research object, the MCNP program of Monte Carlo method is used for modeling and calculation, and the effects of the properties of five SSAs and the moisture content of wet sand on the accident reactivity is classified and quantified to guide the special criticality safety design of the reactor. The results show that the special criticality safety problem of the reactor is solved by the in-reactor use and reasonable arrangement of spectral SSA. The results of this study can provide a useful reference for the criticality safety design of related reactor types.

Numerical Simulation of Flow and Heat Transfer in Shell Side of Lead-Bismuth Eutectic Helical Tube

Shen Cong, Liu Maolong, Liu Limin, Xu Ziyi, Gu Hanyang

2022, 43(S2): 13-18. doi: 10.13832/j.jnpe.2022.S2.0013

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Abstract:
In order to study the flow and heat transfer characteristics in the shell side of LBE Helical-coiled Once-Through Steam Generator (H-OTSG), a numerical method for the simulation of the LBE flow and heat transfer characteristics in helical tube bundles is proposed. Based on available relevant experimental studies, different turbulence models are validated. After validating the reliability of the numerical model, the shell side of LBE H-OTSG is simulated and the influence of velocity and helix angle on heat transfer and flow resistance characteristics are analyzed. The results show that the shell side flow resistance and heat transfer increase with the increase of flow rate and helix angle. An increase in velocity would make the flow fields more uniform. The increase of helix angle affects the vortex distribution and enhances the mixing. This study provides a reference for the study of flow and heat transfer characteristics and design optimization of LBE helical tube.

Study on Radiation Field Distribution and Attenuation Rate of Radioisotope Thermophotovoltaic System

Han Xilong, Lu Mingyu, Yang Aixiang, Shao Jianxiong, Li Ning, Han Chengzhi, Tian Dai, Chen Ximeng, Qiu Jiawen

2022, 43(S2): 19-27. doi: 10.13832/j.jnpe.2022.S2.0019

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Abstract:
For deep space and deep sea exploration missions, the long-period energy supply problem must be solved. Radioisotope thermophotovoltaic (RTPV) cell is one of the important solutions. The research on the performance degradation of RTPV system is the basis of engineering design. The power attenuation of 250 W ²³⁸PuO₂ General-Purpose Heat Source (GPHS) was calculated systematically, and the radiation field distribution inside the thermophotovoltaic system was simulated by Monte Carlo method. Finally, the attenuation rate of RTPV system was obtained by combining relevant foreign experiments. When the cell system has been in service for 20 years, the annual attenuation rate of the system conversion efficiency reduction caused by the thermal power reduction is 0.5%. In the case of oxygen enrichment, the annual attenuation rate of GaSb crystal conversion efficiency caused by neutron irradiation from ²³⁸PuO₂ source is 0.7%. Considering the effects of power attenuation and neutron irradiation, the annual attenuation rate of thermal photovoltaic system is 1.2%.

Atomic Simulation of the Interaction between Dislocation Line and Ferrite/Iron Oxide Interface

Zhu Bida, Yu Xinyang, Li Zheng, He Manru

2022, 43(S2): 28-32. doi: 10.13832/j.jnpe.2022.S2.0028

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Abstract:
Iron oxide is a common surface oxide and internal precipitate of nuclear grade steel containing ferrite phase (such as low alloy steel and ferrite-martensite dual phase steel) under high temperature. A correct understanding of the influence of iron oxide on the micro-deformation mechanism of steel is of great significance to the safety evaluation of advanced nuclear energy systems with higher operating temperatures. In view of this, the effect of temperature on the interaction between edge dislocations in ferrite and ferrite/iron oxide two-phase interface is studied by molecular dynamics method. The calculation results show that in the temperature range of 10~900 K, the edge dislocations can not penetrate the ferrite/oxide interface, but only cause a certain degree of shear deformation between the two phases. With the increase of temperature, the stress concentration near the dislocation-interface contact point increases, and the shear deformation between the interfaces also increases. The above results have certain guiding significance for fracture failure analysis of low alloy steel and ferrite-martensite dual phase steel under high temperature environment.

Numerical Simulation of Reverse Flow in Inverted U-tube Steam Generator under Low Flow Single-phase Condition

Zhang Rui, Wu Qi, Ma Zaiyong, Chu Tao, Zhang Luteng, Tang Yu, Sun Wan, Pan Liangming

2022, 43(S2): 33-39. doi: 10.13832/j.jnpe.2022.S2.0033

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Abstract:
Based on the experimental results of single-phase reverse flow in an inverted U-tube steam generator (UTSG), the primary and secondary sides of the experimental section are modeled proportionally by the method of 3D computational fluid dynamics (CFD). The Realizable k-ε model is selected for turbulence calculation, and the grid independence is verified. By comparing the numerical simulation results with the experimental data, it is found that the calculated values are in good agreement with the experimental values. At the same time, it is found that the critical velocity increases with the increase of the primary side inlet temperature and the decrease of the secondary side temperature; The operating pressure at the primary side has little effect on the critical flow rate.

Electromagnetic Analysis, Check Calculation and Optimization Design of Sodium Electromagnetic Pump Based on Comsol

Xu Shuai, Wang Chong, Chen Shuming, Yu Huajin, Meng Lei, Ai Changjun

2022, 43(S2): 40-46. doi: 10.13832/j.jnpe.2022.S2.0040

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Abstract:
In order to improve the design efficiency of the electromagnetic pump and solve the difficulties in the electromagnetic design and performance analysis of the electromagnetic pump, In this paper, Comsol multiphysics software is used to check and calculate the electromagnetic design part, coupling the electromagnetic module and the fluid module, and the calculation results are consistent with the results of the sodium loop performance experiment, which proves the reliability of the calculation results. In addition, the key parameters affecting the performance of the electromagnetic pump are calculated and analyzed, and an optimization scheme is put forward. By reducing the tooth width, increasing the thickness of sodium flow channel and reducing the thickness of the inner and outer walls of the pump groove, the length of the pump is reduced, the volume is reduced, and the efficiency is increased. The efficiency after optimization is increased by 110%.

Research on RTD Degradation Detection Algorithm Based on Cross-Comparison

Cai Wanrui, Zhang Zhongxiang, Wang Chao, Yin Jichao

2022, 43(S2): 47-52. doi: 10.13832/j.jnpe.2022.S2.0047

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Abstract:
In order to realize the in-situ performance test of resistance temperature detector (RTD) in nuclear power plant, reduce the test error, improve the accuracy of test conclusions and reduce the test cost, a RTD degradation detection algorithm based on cross-comparison is developed, including temperature field instability correction algorithm and temperature inconsistency correction algorithm, and used to downgrade RTD correction algorithm. Based on the overhaul data of a nuclear power unit, the verification results show that the algorithm can reduce the workload and cost of RTD determination during the operation life of the nuclear power plant, improve the accuracy of RTD determination, and provide input for the research of instrument life aging. Therefore, the RTD degradation detection algorithm developed in this study can be used to determine the RTD performance of nuclear power units.

Analysis of Heat Conduction Performance of Heat Pipe Based on Thermal Resistance Network Method

Cui Yinghuan, Xu Jianjun, Xie Tianzhou, Zhou Huihui

2022, 43(S2): 53-59. doi: 10.13832/j.jnpe.2022.S2.0053

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Abstract:
In order to obtain the heat conduction characteristics such as heat pipe temperature and transmission power, a 2D heat pipe thermal resistance network model considering steam cavity heat transfer is established based on the thermal resistance equivalence theory, and the transient and steady-state heat transfer performance of the heat pipe are obtained. The literature and the experimental data of this paper are compared and verified. The maximum relative error of the average temperature in the adiabatic section is 7.6%. The results show that the model has a high accuracy for the experimental data of sodium working medium heat pipe in literature and hydraulic working medium heat pipe in this paper; The temperature uniformity of each axial section of the heat pipe is good, and the radial thermal resistance of the wick is the main factor affecting the thermal-conduction resistance of the heat pipe; The working temperature and transmission power of the heat pipe change with the cold source parameters in the opposite trend, and the heat transfer of the heat pipe is an adaptive dynamic regulation process. Therefore, the thermal resistance network model can be used as a tool for experimental analysis and heat pipe design. At the same time, it can further expand the research scope of working conditions, obtain the influence rules of boundary conditions on heat pipe heat transfer process, and further provide reference for heat pipe design in multiple application scenarios.

Study on Application of Predictive Maintenance Technology in Nuclear Power Plant

Shang Xianhe, Zeng Chun, Li Wei

2022, 43(S2): 60-66. doi: 10.13832/j.jnpe.2022.S2.0060

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Abstract:
In order to ensure the safe, stable and economic operation of nuclear power plant units, nuclear power enterprises have gradually introduced predictive maintenance (PdM) based on equipment status. Based on the production practice of a nuclear power plant, this paper discusses the development and application of PdM technology and the challenges faced in the practical application of PdM technology in nuclear power plants, and points out that in consideration of the complexity and particularity of nuclear equipment, when applying PdM technology, different maintenance strategies shall be reasonably selected according to the classification of equipment, the failure mechanism of equipment and the maturity of PdM technology itself; at the same time, attention shall also be paid to the collection and analysis of equipment data information in order to optimize predictive maintenance technology, so as to obtain satisfactory equipment management results.

Study on Thermal Deformation Behavior and Mechanism of Difficult-to-process Boron-containing Stainless Steel

Li Yongwang, He Xueyi, Wang Xinmin, Guo Zhen, Wu Yu, Liu Haitao

2022, 43(S2): 67-73. doi: 10.13832/j.jnpe.2022.S2.0067

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Abstract:
The solid solubility of boron in steel is low, and a large number of coarse, hard and brittle eutectic borides are easily formed in the solidification process of Boron-containing stainless steel prepared by fusion casting, which leads to poor hot processibility of its billets. In order to solve this problem, using 18.5Cr-14.0Ni-2.1B stainless steel as the research material, the single-pass hot compression experiment was carried out on a thermal simulation machine, the flow stress constitutive equation was established, the hot processing map was constructed, the reasonable processing window of the stainless steel was defined, and the thermal deformation behavior and mechanism of boron-containing stainless steel were revealed. The results show that the apparent thermal deformation activation energy Q and stress exponent n of 18.5Cr-14.0Ni-2.1B stainless steel are 501.08 kJ·mol⁻¹ and 8.13 respectively; In the process of hot compression, eutectic borides are mainly broken, rotated and induced to form a large number of micropores, most of which can be filled by the plastic flowing austenitic matrix; The larger the strain of the hot compression process is, the smaller the processing window is. When the 18.5Cr-14.0Ni-2.1B boron-containing stainless steel is deformed at 1000~1100℃ and at the rate of 0.01~1.0 s⁻¹, it shows good processibility.

Numerical Simulation on Milling Process of Ni-based Alloy Welds and Optimization Analysis of Milling Cutter

Xu Shaofeng, Yang Biao, Ye Yihai, Zhang Jiahao, Zhang Yilin, Chen Jiahao

2022, 43(S2): 74-81. doi: 10.13832/j.jnpe.2022.S2.0074

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In order to optimize the milling process and cutter performance of nickel-base alloy welds, the finite element model of nickel-base alloy weld milling was established based on Abaqus simulation software, and the milling process simulation calculation analysis and cutter failure research were carried out. The effects of different cutting parameters and cutter parameters on cutting temperature and cutting force were analyzed. Combined with the calculation results, the orthogonal test of cutting parameters and the optimization analysis of cutter parameters were completed. The results show that smaller cutting force and cutting temperature can be obtained with smaller cutting speed, small feed rate and increased air cooling, and the cutting force and cutting temperature can be effectively reduced by adjusting the cutter front angle and the blunt radius of the cutter tip. The research results can provide reference for the optimization of cutting parameters and cutter parameters in the milling process.

Study on Tangential Fretting Wear Behavior of Zirconium Alloy Cladding at High Temperature

Ren Quan yao, Pu Zeng ping, Jiao Yongjun, Zheng Meiyin, Chen Ping, Han Yuanji, Liu Menglong, Zhuang Wenhua, Guo Xianglong, Zhang Lefu

2022, 43(S2): 82-87. doi: 10.13832/j.jnpe.2022.S2.0082

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Abstract:
During the service of the nuclear fuel assembly, the fretting wear between the grid spacer holding structure and the fuel rod is the first major factor leading to the damage of the fuel rod cladding, accounting for about 54.8% of the fuel rod cladding failure. According to the tangential fretting wear behavior of zirconium alloy cladding tubes with different holding structures, the experimental study on wear under high temperature and high pressure hydrochemical environment is carried out. The key parameters, such as cladding tube wear scar morphology, wear volume and wear depth of curved and planar holding structures under different holding forces are compared and analyzed. The results show that the wear of the curved holding structure on the fuel rod cladding is mainly abrasive wear, lamellar shedding and “ploughing” effect, while the planar structure is dominated by “ploughing” effect and lamellar shedding, and the abrasive wear is less. In addition, under the same conditions, the maximum wear depth of the curved holding structure is larger than that of the planar holding structure.

Effect of Carbon Source on Boron Loss in Preparation of Zirconium Diboride by Carbothermal Reduction

Yu Chong, Leng Ke, Wang Yi, Zeng Qiang, Tang Yan, Liu Zhe

2022, 43(S2): 88-93. doi: 10.13832/j.jnpe.2022.S2.0088

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Abstract:
In order to solve the problem of easy loss of boron in the preparation of nuclear grade ZrB₂ target used in the third-generation nuclear power reactor AP1000, ZrO₂ powder was prepared by carbothermal reduction method using ZrO₂, B₄C and C as raw materials. The effects of carbon source and carbon addition on boron loss during high temperature synthesis were studied. The phase composition, micro morphology and element content of the samples were analyzed by X-ray diffraction (XRD), scanning electron microscopy (SEM) and chemical titrator. The loss of boron content in the process of powder synthesis was calculated according to the boron content before and after synthesis and the principle of mass conservation. The results show that the loss of boron can be caused by the preparation of ZrB₂ powder by carbothermal reduction; Compared with graphite, using carbon black as the carbon source with appropriate excess can effectively reduce the loss of boron in the synthesis process; When B₄C exceeds by 5% and carbon exceeds by 20%, the boron loss rate in the synthesis process is the lowest 4.4%, and the particle size of ZrB₂ powder is uniform, the particle size is 1-2 1~2 μm, and the purity is high.

Numerical Study on Heat Transfer Enhancement of Modified Wall in U-tube Steam Generator

Yuan Junjie, Liu Li, Gu Hanyang

2022, 43(S2): 94-99. doi: 10.13832/j.jnpe.2022.S2.0094

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Abstract:
In order to achieve better heat transfer effect and reduce the possibility of pipe rupture caused by heat transfer deterioration, based on the improved lattice Boltzmann method (LBM) phase transition heat transfer model, the bubble dynamics behavior and wall heat transfer performance of working medium in U-shaped heat transfer tube of steam generator were studied numerically. The results show that compared with the pure hydrophilic wall, the hydrophobic zone promotes bubble nucleation, and the wall with mixed wettability can significantly increase the boiling heat transfer. The wall with mixed wettability can increase the critical heat flux (CHF) and delay the occurrence of CHF point, which can effectively reduce the possibility of heat transfer deterioration. Generally speaking, the distance and number of hydrophobic points determine the growth state of bubbles, the best boiling heat transfer performance can be obtained if there is an optimal setting of hydrophobic points, and the ideal heat transfer wall can be obtained by selecting appropriate distance and number of hydrophobic points.

Preliminary Study on Conceptual Design of Lead-based Fully Ceramic Microencapsulated Dispersion Fuel Core

Lou Lei, Wang Lianjie, Peng Xingjie, Zhao Chen, Zhang Bin, Zhou Bingyan, Zhou Nan, Hu Yuying, Wang Xingbo, Zhao Zifan

2022, 43(S2): 100-103. doi: 10.13832/j.jnpe.2022.S2.0100

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Abstract:
In order to make full use of the accident resistance characteristics of fully ceramic microencapsulated (FCM) dispersion fuel and further improve the safety of lead-based reactor, FCM fuel is applied to lead-based coolant reactor, and the preliminary conceptual design of lead-based FCM core is presented, and the fuel loading, fuel utilization, energy spectrum and reactivity of the FCM core are compared with the traditional lead-based UO₂ fuel core. The comparison results show that FCM has a little moderating effect on core energy spectrum, and high enrichment UO₂ fuel core is needed to ensure that the core ²³⁵U loading meets the energy output requirements. The ²³⁵U loading of the core using FCM is lower than that of UO₂ core, and the fuel utilization rate is further improved. Finally, the power flattening optimization of lead-based FCM core layout is carried out, and the core power is flattened by radial FCM phase volume partition. The calculation results show that the core power peak factor (F_Q) is reduced from 2.43 to 1.93, and the core nuclear enthalpy rise factor (F_DH) is reduced from 1.79 to 1.33.

Study on the Diffusion and Release Characteristics of Fission Products from TRISO Coated Fuel Particles

Zheng Junqiang, He Yanan, Zhang Jing, Wu Yingwei, Tian Wenxi, Su Guanghui, Qiu Suizheng

2022, 43(S2): 104-110. doi: 10.13832/j.jnpe.2022.S2.0104

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To analyze the diffusion and release characteristics of TRISO-coated fuel particles fission products, the Fick diffusion model of fission products of TRISO-coated fuel particles under irradiation-heat-force coupling is established and verified by the IAEA CRP-6 benchmark task. Then the performance of TRISO-coated fuel particles under typical working conditions of a high-temperature gas-cooled reactor is also analyzed by using the model. At the same time, the release characteristics of fission products at different temperatures and particle powers of TRISO-coated fuel particle are analyzed considering the recoil effect and thermal diffusion effect of fission products. The results show that high temperature causes the loss of the fission product containment capacity of TRISO-coated fuel particles, and the increase of power has less impact on the release of fission products.

Study on On-the-fly Cross-Section Treatment and Burnup Calculation of Space Nuclear Reactor Based on Monte Carlo Method

Li Rui, Liu Shichang, Che Rui, Lu Di, Wang Lianjie, Wang Zhenyu, Chen Yixue

2022, 43(S2): 111-117. doi: 10.13832/j.jnpe.2022.S2.0111

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To solve the problem of on-the-fly cross-section treatment in the calculation of the Monte Carlo (MC) code, the target motion sampling method and the improved Gauss-Hermite method are used to calculate the effective multiplication factor ( k_eff) at different fuel temperatures for the resolvable resonance energy region. Two cross-section treatment methods are applied to the kilowatt-class heat pipe space nuclear reactor model and compared with the calculation results of keff using the accurate temperature cross-section library. The results show that for k eff at different temperatures, the absolute errors of the two methods are within ± 3 times the relative combined statistical error. The calculation time of the target motion sampling method and the improved Gauss-Hermite method increases by 45% and 9%, respectively. To solve the burnup calculation problem in space reactors, the internal coupling burnup calculation function of the Reactor Monte Carlo code (RMC) is applied to the megawatt-class heat pipe space nuclear reactor model and the prism-type high-temperature gas-cooled reactor model, and compared with typical MC code Serpent and MCNP. The results show that the maximum error of k_effis less than 0.2% and 0.3% at each burnup step. Two studies preliminarily verify the correctness of RMC's on-the-fly cross-section treatment and burnup calculation of space nuclear reactors.

Core Design and Analysis of Physical and Thermal Characteristics of Lunar-based Nuclear Power Source

Ning Kewei, Song Shuai, Zhao Fulong, Xie lin, Tan Sichao

2022, 43(S2): 118-124. doi: 10.13832/j.jnpe.2022.S2.0118

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Nuclear energy has the characteristics of long endurance and long life, which can provide reliable energy guarantee for lunar exploration. In this paper, the lunar-based 300kWt thermionic reactor scheme is proposed. The core physical characteristics are analyzed by Monte Carlo method, and the reactor thermal safety is studied by numerical simulation through the single channel flow and heat exchange calculation of the core. The calculation results show that the scheme can achieve 10 years’ lifetime and has sufficient shutdown depth, but there is positive temperature feedback in some temperature intervals, and it needs to rely on the fine control to meet the reactor safety requirements. In the lunar microgravity environment, when the coolant flow rate is not less than 0.9m/s, the hottest channel of the core does not exceed the thermal safety limit.

Thermal-Hydraulic Investigation of LBE Cooled Wire-Wrapped Fuel Bundle Based on Entropy Generation Analysis

Zhang Dong, Zhang Haochun, Wang Qi, Sun Wenbo

2022, 43(S2): 125-130. doi: 10.13832/j.jnpe.2022.S2.0125

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In order to analyze the thermal hydraulic characteristics of LBE cooled fast reactor fuel assemblies from the point of view of design and operation, based on the finite volume method, 19 LBE cooled fuel rod bundles with wire-wrapped structures are numerically simulated. The flow and heat transfer characteristics of fuel assemblies under different mass flow rates and thermal power are analyzed. The entropy generation characteristics and thermodynamic irreversibility of coolant under different working conditions are studied by entropy generation analysis method. The results show that the secondary flow and the distribution of entropy generation show periodic changes in the axial direction; the inlet velocity is the main factor affecting the secondary flow and entropy generation distribution; On the premise of ensuring structural safety, increasing coolant flow rate appropriately is conducive to improving the thermal economic performance of the coolant.

Analysis of Burnup Characteristics of Chloride Salt Fast Reactor under Pre-breeding Scheme

He Liaoyuan, Zou Yang, Yan Rui

2022, 43(S2): 131-136. doi: 10.13832/j.jnpe.2022.S2.0131

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In order to analyze and evaluate the Th-U cycle scheme of starting reactor with low enriched uranium (LEU), Pu and transuranium (TRU) in fast neutron chloride salt reactor, the concepts of regeneration ratio and replacement ratio are defined to compare the conversion information of key nuclides in the Th-U cycle reaction chain. The burnup analysis of pre-breeding scheme under three different start-up fuels is carried out by using TMCBurnup program. The analysis results show that the under the pre-breeding cycle scheme, the production of ²³³U can be quickly completed by using three fuels for start-up, in which the TRU start-up transition mode has a larger replacement ratio, the LEU start-up transition mode has a larger ²³³U regeneration ratio, and the core always has a large negative temperature feedback coefficient in the whole operation cycle.

Study on Operating Characteristics and Water Hammer Phenomenon of Lead-Bismuth Eutectic Commutator with Enclosed Structure

Fan Xukai, Zhu Yucheng, Peng Tianji, Tian Wangsheng, Tang Yanze, Fan Deliang

2022, 43(S2): 137-142. doi: 10.13832/j.jnpe.2022.S2.0137

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In order to study the flow stability and applicability of the commutator with enclosed structure in the process of liquid lead-bismuth eutectic (LBE) medium commutation, FLUENT software combined with overlapping grid method is used to calculate and analyze the pressure and velocity fluctuations upstream of the commutator with enclosed structure in the form of three-way switching valve during commutation. The results show that in the process of commutation, the water hammer is produced when the valve element starts and stops, and the water hammer wave propagates in the pipe at the medium sound velocity, thus affecting the distribution of pressure and velocity in the pipe. The peak value of the water hammer wave is proportional to the medium density and the movement speed of the valve element, and the frequency of the water hammer wave is not affected by the movement speed of the valve element. Because the peak value of the water hammer wave is large, it is difficult for the upstream pipe of this type of commutator with enclosed structure to maintain a steady flow and threaten the safety of the device during commutation operation under LBE, so it is not suitable for LBE flow calibration device.

Method Study and Application for Neutronics Optimization of Solid Tritium Breeding Blanket of Fusion Reactor

Qu Shen, Cao Qixiang, Wu Xinghua, Yin Miao, Zhao Fengchao, Wang Xueren, Duan Xuru, Wang Xiaoyu

2022, 43(S2): 143-149. doi: 10.13832/j.jnpe.2022.S2.0143

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This paper aims to improve the tritium breeding performance of the Tritium Breeding Blanket (TBB) of fusion reactors to better meet the tritium self-sufficiency. First, based on neutronics perturbation theory and simulated annealing algorithm, a new algorithm and program for neutronics optimization of TBB are developed. Second, the helium cooled solid blanket of China Fusion Engineering Test Reactor（CFETR） is selected to complete the demonstration application of the whole reactor neutronics performance optimization. Finally, the 3D finite element check of thermal, fluid and structure is carried out for the optimized blanket scheme. The results show that: ① The optimization algorithm proposed in this paper has better optimization effect and higher optimization efficiency than the traditional blanket neutronics optimization algorithm; ② The intelligent optimization program developed in this paper can better meet the needs of neutronics optimization and design of the fusion reactor blanket, and provide the theoretical basis of algorithm and program support for the blanket design.

CFD Study on Flow and Heat Transfer Characteristics of Counter-flow Straight Channel Printed Circuit Plate Heat Exchanger

Mao Junjun, Yang Xiaoyong, Zhao Gang, Peng Wei

2022, 43(S2): 150-157. doi: 10.13832/j.jnpe.2022.S2.0150

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Printed circuit heat exchangers (PCHEs) have a good application prospect in the Helium closed Brayton cycle of high temperature gas cooled reactor because of its high compactness, high temperature and high pressure resistance and high heat transfer efficiency. In this paper, the flow and heat transfer performance of counter-flow straight channel PCHE helium regenerator is studied by using computational fluid dynamics (CFD) numerical simulation. The effect of geometric parameters on helium flow and heat transfer in straight channel PCHE is analyzed by changing the channel horizontal spacing, vertical spacing and flow channel radius. The calculation results show that the vertical spacing only has an effect on the heat transfer performance, while the horizontal spacing has no obvious effect, and the flow channel radius has an effect on the flow and heat transfer performance of PCHE. The heat transfer performance of PCHE is improved by reducing the channel vertical spacing; When the inlet mass flow rate is kept constant, if the channel radius is reduced, the convective heat transfer coefficient and heat transfer performance are enhanced, but the pressure drop along the channel increases. In addition, the vertical or horizontal multi-channel PCHE model is used to compare the temperature and pressure drop distribution along different channels to determine the effect of vertical and horizontal positions on the thermal and hydraulic performance of PCHE. By comparing the calculation results of the multi-channel model with that of a single heat exchange unit, it is shown that a single heat exchange unit with periodic boundary conditions can more accurately simulate most of the flow channels of the complete PCHE model.

Fluid-structure Interaction Simulation and Data-driven Modeling of Tube Bundle Based on OpenFOAM

Feng Zhipeng, Zhang Yixiong, Huang Xuan, Liu Shuai, Qi Huanhuan, Cai Fengchun

2022, 43(S2): 158-164. doi: 10.13832/j.jnpe.2022.S2.0158

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In order to realize the application of open-source tool OpenFOAM in the prediction of tube bundle fluid-structure interaction, in view of the lack of comprehensive benchmark cases for large eddy simulation verification in OpenFOAM, the lack of parameter identification methods and data-driven modeling methods based on OpenFOAM simulation data, this study first quantitatively compares the performance of large eddy simulation in OpenFOAM by studying benchmark problems, focusing on the effects of statistical time length, size and shape of calculation domain, meshing method, and wall function on the results, and verifies the numerical results with experimental data to obtain a reasonable flow field analysis model; Then, this study couples the motion equation with the flow field calculation, solves the unsteady Navier-Stokes (uRANS) equation with moving boundary, realizes the fluid-structure interaction simulation of the tube bundle, and successfully captures the fluid-structure interaction characteristics of the tube bundle. Taking the flow-cell model as an example, this study realizes the key parameter identification and data-driven modeling. The results show that at least 180 vortex shedding cycles are required to achieve statistical convergence in large eddy simulation; The lift and recirculation length are sensitive to the grid resolution, while the vortex shedding frequency and cylinder surface pressure are sensitive to the calculation domain; For the statistical distribution of the wake zone, the effect of grid resolution is more significant, the effect of the shape of the calculation domain can be ignored, and the critical velocity calculated by data-driven modeling is in good agreement with the experimental data.

Numerical Simulation Study on Segmented Welding Deformation of Main Pipe of Pressure Vessel

Yu Mingda, Zhang Liping, Shao Xuejiao, Jiang Lu, Li Hui, Liu Zhengu, Pu zhuo

2022, 43(S2): 165-170. doi: 10.13832/j.jnpe.2022.S2.0165

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Abstract:
For the excessive radial deformation problem of the reactor pressure vessel during the welding process, based on SYSWELD, the numerical simulation analysis of the multi-pass welding of a main pipe is conducted, and the effects of three sub-sectional welding schemes on the welding residual stresses and welding deformation are investigated. The results show that, the numerical results are in good agreements in the experimental data, so the numerical approach involved is validated. The sub-sectional schemes can effectively reduce the radial displacements of the pressure vessel main pipe during the welding process. Among the three welding schemes, the sub-sectional up-to-down single-side scheme shows the best effect on the welding radial deformation suppression.

Sensitivity Analysis for Dynamic Response of Reactor Coolant System Based on OPTIMUS

Yuan Yanli, Zhang Yixiong, Ye Xianhui, Wang Bihao, Li Bingjin, Yang Kang

2022, 43(S2): 171-176. doi: 10.13832/j.jnpe.2022.S2.0171

Abstract(102) HTML (99) PDF(20)

Abstract:
In order to obtain the design parameters that have a greater impact on the dynamic response of the reactor coolant system, shorten the design cycle, and improve the design efficiency, this paper uses the support stiffness and support clearance of the steam generator (SG) in the reactor coolant system (RCS) as input variables, and uses OPTIMUS integration platform to carry out the sensitivity analysis of the system dynamic response to the input variables under seismic conditions. The analysis shows that the bending moment of the inlet nozzle of the main pump is more sensitive to the clearance of the lower horizontal support of SG. The relevant sensitivity analysis method and analysis process can be extended to other systems and equipment of nuclear power plant, providing quantitative analysis means and data support for reliability analysis parameter selection and optimization design improvement of reactor coolant system.

Calculation of Stress Intensity Factor for Surface Cracks in FeCrAl Cladding under Force-Irradiation Coupling

Zhu Bida, Shi Kaikai, Zheng Bin

2022, 43(S2): 177-181. doi: 10.13832/j.jnpe.2022.S2.0177

Abstract(198) HTML (61) PDF(32)

Abstract:
In order to study the resistance of Ferritic FeCrAl alloy cladding to fracture failure under high radiation dose, the crack tip stress intensity factor of axial semi-elliptical surface crack of FeCrAl cladding tube under internal pressure and radiation swelling stress coupling load is calculated by finite element method combined with crack block analysis technique in Zencrack software. The change rule of stress intensity factor at crack tip with different depth on inner and outer surfaces under different radiation swelling strain distribution is obtained, and the influence of radiation swelling effect is analyzed. The results show that compared with the uniform radiation swelling strain distribution, when the radiation swelling strain of the cladding tube has a linear gradient in the radial direction, the crack tip stress intensity factor of the axial semi-elliptical crack on the inner surface decreases significantly. On the other hand, the crack tip stress intensity factor of the axial semi-elliptical crack on the outer surface increases significantly.

Research on Design Method of HPR1000 Reactor Vessel Internals

Li Hao, Li Yan, He Peifeng, Yu Zhiwei, Hu Chaowei, Wang Qingtian, Xia Xin, Zhao Wei

2022, 43(S2): 182-188. doi: 10.13832/j.jnpe.2022.S2.0182

Abstract(264) HTML (77) PDF(63)

Abstract:
The forward design method of HPR1000 reactor vessel internals (RVI) is studied to respond to the new design requirements of HPR1000 third-generation nuclear power reactor and standardize the structural design of HPR1000 RVI. RVI is complex equipment at the system level, and its structural design affects many aspects of reactor performance. Through the demand analysis of RVI functions, the causal relationship among various disciplines is clarified, and the functional requirements items that RVI design needs to meet are established. At the same time, based on the functional requirements of RVI and the empirical feedback of reactor type at home and abroad, the quantitative evaluation criteria for important functional requirements items are established, and the quantitative indicators are used to standardize the RVI design. Through the study of RVI design method, a total of 72 functional requirements items in 10 aspects of HPR1000 RVI are established. At the same time, according to the analysis and calculation, the quantitative evaluation criteria of each analysis are determined, and clear indicators are put forward for each analysis and calculation. This method realizes the forward design of RVI, covers its full life cycle functional requirements, and ensures the reliability and safety of RVI design.

Research on Technology of Isolated Magnetoresistive Multipolar Resolver

Tang Yuan, Wu Hao, Yan Dapeng, Huang Siyu, Li Qingzhao, Liu Yanting, Tang Jiankai, Fu Guozhong, Zhang Zhiqiang, Zhang Rui, Cui Zongze, Song Liwei

2022, 43(S2): 189-195. doi: 10.13832/j.jnpe.2022.S2.0189

Abstract(212) HTML (82) PDF(20)

Abstract:
In order to meet the demand of angle measurement in high temperature and high pressure severe environment, a stator and rotor isolated magnetoresistive multipolar resolver which can be used in harsh environment is studied. A magnetic conducting stainless steel shielding shell is installed between the stator and rotor of such isolated magnetoresistive multipolar resolver. For this type of isolated magnetoresistive multipolar resolver, the magnetic circuit model of the magnetoresistive multipolar resolver with shielding shell is established, the design basis of such multipolar resolver is given, and the problem of the shielding shell taking away the main magnetic flux is analyzed and simulated by finite element method. The abnormal phenomenon of output signal of isolated magnetoresistive multipolar resolver is analyzed, and the compensation methods of bias and amplitude anomaly are designed. The results of simulation analysis show that the isolated magnetoresistive multipolar resolver can obviously suppress the fluctuation of position solution caused by abnormal signals.

Numerical Simulation of Oxidation Corrosion behavior of F-M Steel in LBE Environment Based on Lattice Boltzmann

Wu Jiayue, Luo Ying, Du Hua, Wang Liubing, Wu Bingjie, Zhu Mingdong

2022, 43(S2): 196-201. doi: 10.13832/j.jnpe.2022.S2.0196

Abstract(249) HTML (141) PDF(25)

Abstract:
Liquid lead-bismuth eutectic (LBE), which is the coolant medium of lead-cooled fast reactor, seriously corrodes the structural materials of the reactor. Ferritic-martensitic steel (F-M steel), as a candidate material for reactor structure, undergoes oxidation corrosion in oxygen controlled LBE, forming a typical double-layer oxide film structure. In order to design a highly reliable reactor structure and predict the material life of F-M steel in LBE environment, based on lattice Boltzmann (LBM) method, this paper simulates the corrosion phenomena such as multi-component transport, oxidation reaction, solid liquid phase transition, etc. in the process of oxidation corrosion, and establishes the oxidation corrosion model of stainless steel in LBE. The calculated simulation results are in good agreement with the experimental data, and the established model can explain the role of diffusion and reaction in the oxidation process. The lattice Boltzmann model developed can be used to study the growth of mesoscopic oxide films.

Study on High-temperature Creep for Lower Head of HPR1000 Reactor Pressure Vessel

Yang Licai, Qiu Tian, Yang Zhihai, Yin Qiwei

2022, 43(S2): 202-207. doi: 10.13832/j.jnpe.2022.S2.0202

Abstract(170) HTML (46) PDF(34)

Abstract:
High-temperature creep is main failure mode of HPR1000 RPV lower dome under severe accident. To accurately study high-temperature creep of HPR1000 RPV lower dome made of domestic 16MND5 forging and assure structure integrity of RPV lower dome under severe accident, high-temperature creep of HPR1000 RPV lower dome is studied systematically in this paper by combining numerical simulation and theoretical analysis based on the high-temperature creep test data. Firstly, the constitutive model of RPV lower dome material is established. Adopting ANSYS software, the numerical simulation of the high-temperature creep of the lower dome under the action of high temperature and internal pressure is performed, and creep strain and stress distributions for lower dome are obtained. Furthermore, the high-temperature creep problem of RPV lower dome is theoretically studied for the first time. The research results indicate, high-temperature creep of RPV lower dome mainly occurs in the zone whose temperature is higher than 450℃; Under severe accident, HPR1000 RPV doesn’t fail due to high-temperature creep ; Plasticity failure zone will enlarge with the increase of internal pressure; The theoretical analysis results of steady creep are consistent with the numerical simulation ones, and deeply reveal the layered failure phenomenon of RPV lower dome.

Study On Hydraulic Properties of Bottom Nozzle of CF3 Fuel Assembly

Feng Linna, Chen Jie, Su Min, Pu Zengping

2022, 43(S2): 208-212. doi: 10.13832/j.jnpe.2022.S2.0208

Abstract(234) HTML (71) PDF(36)

Abstract:
Bottom nozzle is an important component of CF3 fuel assembly, with the important functions of filtering foreign matters, circulating coolant and supporting the fuel assembly. In this paper, single-phase computational fluid dynamics (CFD) technology and hydraulic test methods are respectively used to study the hydraulic properties of CF3 fuel assembly in view of the flow function of the bottom nozzle. The calculation and test results show that the resistance coefficient of the bottom nozzle of CF3 fuel assembly is 5% different from that of AFA3G, which meets the requirements of hydraulic design of CF3 fuel assembly in terms of coolant fluidity.

Research on Predictor-Corrector Method Based on Prediction of Reaction Rates

Wen Xingjian, Tian Chao, Tang Songqian, Zhai Zian, Miao Jianxin, Cao Liangzhi, Liu Zhouyu

2022, 43(S2): 213-219. doi: 10.13832/j.jnpe.2022.S2.0213

Abstract(214) HTML (120) PDF(22)

Abstract:
Traditional predictor-corrector method revises the burnup calculation results by means of the nuclide inventory of the average predictor step and the corrector step, two neutronics calculations are required at each burn-up step. In order to reduce the time of transport-burnup coupling calculation, the traditional predictor-corrector method is analyzed in this work and the predictor-corrector method based on the prediction of the reaction rates is proposed in this work. This method makes use of the characteristics of the smooth variation of the multi-group neutron spectrum difference between the predictor step and the correction step with burnup in the traditional predictor-corrector method, Lagrangian interpolation method is used to describe the variation characteristics of regional and burnup time related neutron spectrum deviation with burnup time, and analytical interpolation formula is obtained. Under the condition that the calculation times of the predictor steps remain unchanged, the reaction rate of some corrector steps is predicted directly so as to avoid the neutronics calculation of such corrector steps. This study takes the traditional predictor-corrector method as reference, and verifies the proposed predictor-corrector method based on reaction rate prediction through VERA and JAEA burnup benchmark tasks. The numerical results show that the predictor-corrector method based on reaction rate prediction can reduce the neutronics calculation time by about 20% on the basis of ensuring the rod power distribution and the accuracy of the main nuclide components. Therefore, the predictor-corrector method based on reaction rate prediction established in this study can be used for efficient transport-burnup coupling calculation.

Researh on Direct Current Feedback Control Technology for Inverter Power

Chen Meiyuan, Yu Haitao, Peng Renyong, He liang, Wang Chunlei, Liu yiyi, Zhang Jianjian

2022, 43(S2): 220-223. doi: 10.13832/j.jnpe.2022.S2.0220

Abstract(112) HTML (27) PDF(14)

Abstract:
Inverter power is the important equipment to realize reactivity control of nuclear reactor. The existing inverter power can not keep the actual DC operating current constant when the power supply fluctuates and the operating frequency changes. In order to make the actual DC operating current constant, this paper studies a DC feedback control technology of inverter power, which takes the actual DC operating current as feedback, uses proportional-integral regulator for feedback control, and uses space vector pulse width modulation (SVPWM) algorithm to modulate the given operating frequency and regulator output into pulse width modulation (PWM) signals that drive the inverter bridge. The feasibility of this method is demonstrated by simulation, and a feedback inverter power is developed. The effectiveness of this technology is verified by experiments.

Research for Pulse Counting Interference Recognition Technique Based on Average Value of Power Spectrum

Gao Zhiyu, Luo Tingfang, Bao Chao, Zhu Hongliang, Yuan Hang

2022, 43(S2): 224-227. doi: 10.13832/j.jnpe.2022.S2.0224

Abstract(125) HTML (37) PDF(9)

Abstract:
Aiming at the interference of pulse counting in the source range of the Ex-core nuclear measurement system, a method to identify the interference of pulse counting using the average value of power spectrum of pulse counting rate as the decision value is proposed. The decision value of the average value of power spectrum is calculated according to the real reactor data, and three values are set according to the pulse counting rate size. The simulation method is used to verify the effect of interference identification, and the interference counting is superimposed on the pulse counting rate data of the real reactor. The simulation results show that the method can identify the interference signal with a duration of 1 s and an interference intensity (the ratio of interference to pulse counting rate) above 5%.

Research and Design of Diversity Protection System for Small Modular Reactor

Zhu Pan, Xi Mengmeng, Xu Dongfang, Qiu Zhifang, Liu Hongchun, Zhong Sijie, Dang Gaojian

2022, 43(S2): 228-233. doi: 10.13832/j.jnpe.2022.S2.0228

Abstract(121) HTML (39) PDF(22)

Abstract:
According to the characteristics of small modular reactor, the key technologies of diversity protection system (RDA) are studied and analyzed. The design method of diversity protection signal based on the combination of probability theory and certainty theory is adopted for the first time, and the determination of protection signal setting value and delay time is deeply studied, which realizes the minimization of protection signal setting and the maximization of protection function; At the same time, with the comprehensive application of diversity and independent design technology in RDA design, a system architecture with high reliability and low misoperation rate is constructed, which significantly improves the safety and economy of small modular reactor, meets the requirements of defense in depth and diversity of nuclear power plant, and the relevant technical indicators have reached the international advanced level.

Neutronics Analysis on MOX assemblies for HPR1000 Cores

Liu Kun, Liu Tongxian, Jiang Zhumin

2022, 43(S2): 234-238. doi: 10.13832/j.jnpe.2022.S2.0234

Abstract(219) HTML (71) PDF(34)

Abstract:
In combination with the development needs of nuclear energy in China, the neutronics characteristics of mixed oxide (MOX) fuel assemblies are studied using SCIENCE software with HPR1000 reactor as the research object, which lays a theoretical foundation for the subsequent large-scale loading of MOX fuel assemblies in HPR1000 unit and the realization of closed fuel cycle. Numerical analysis is carried out on the influencing factors such as initial enrichment, discharge burnup and core operation parameters of CF3 fuel assembly, and the effects of assembly layout, plutonium vector, total plutonium loading and matrix uranium composition on key neutronics parameters such as reactivity of MOX fuel assembly are evaluated. The numerical results show that when MOX modules are arranged around, it is conducive to achieving better reactivity control capability, reducing reactivity control pressure, making better use of existing core design conditions, and realizing the utilization of plutonium resources. When the fission nuclide composition in the fuel rod is increased, the energy spectrum of the assembly hardens, and the reactivity feedback ability and reactivity control capability are further weakened. For the design scheme of HPR1000 reactor core, additional reactivity control means may be required to realize the reactivity control of MOX fuel assembly during burnup.

2022 Vol. 43, No. S2