核动力工程

Status and Progress of the Multi-Physics Coupling and Multi-Scale Coupling Research for Numerical Reactors

Liu Xiaojing, Xie Qiuxia, Chai Xiang, Cheng Xu

2021, 42(5): 1-7. doi: 10.13832/j.jnpe.2021.05.0001

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Abstract:
This paper explains the basic concept of numerical reactors, and studies in detail the international research and development (R&D) projects for numerical reactors, such as the Consortium for Advanced Simulation of Light Water Reactors (CASL), European Nuclear Reactor Simulation (NURESIM) platform, and Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. It also presents a further study on and a summary of the current research status of multi-physics coupling and multi-scale coupling technology in China and other countries. In combination with the research status, this paper indicates that the focus of the numerical reactor technology development rests on the multi-physics coupling mechanism under the combined action of material corrosion, flow heat transfer and neutronics, and the development of high-fidelity coupling code based on unified grid solution.

Comparison of Subcooled Flow Boiling in a Full-Length 5×5 Rod Bundle between Uniform and Non-Uniform-Axial Power Distribution

Zhang Junyi, Yan Xiao

2021, 42(5): 8-14. doi: 10.13832/j.jnpe.2021.05.0008

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Abstract:
The structure of PWR fuel assembly is in form of square rod bundle. In this study, Computational Fluid Dynamics(CFD) method is verified to compare thermal-hydraulic(T-H) characteristics between Uniform-Axial Power Distribution(U-APD) and Non-Uniform-Axial Power Distribution(Non-U-APD) under subcooled flow boiling condition. It is shown that CFD method has the ability to achieve a goog agreement on the void fraction prediction while reasonalbe wall boiling model, interface force model and bubble size distribution model are implemented. The comparison shows that the onset of significant void point of Non-U-APD bundle appear in advance and the averaged void of Non-U-APD along the axial direction has a more higher increasing rate rather than that of the U-APD. At the end of the heated length, the corner channel-avereged void of Non-U-APD is higher than that of the U-APD while the central channels are nearly the same, although the same inlet condition and heated power is applied in both the simulation.At the downstream of 5th and 6th Mixing Vane Grid(MVG), the liquid mass flux of corner and central channel in Non-U-APD are lower than that of U-APD due to the phase change.

Code Development and Engineering Validation of PWR Fuel Management Software Bamboo-C

Wan Chenghui, Li Yunzhao, Zheng Youqi, Liu Zhouyu, Zu Tiejun, Cao Liangzhi, Wu Hongchun, Shen Wei

2021, 42(5): 15-22. doi: 10.13832/j.jnpe.2021.05.0015

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Abstract:
Based on the conventional “two-step” scheme for the PWR fuel management, an advanced PWR fuel management software Bamboo-C has been developed by the most advanced methodologies in the reactor-physics field. Bamboo-C consists of three main functional codes: LOCUST code for the heterogeneous modeling and simulation and homogenization calculation of 2D assemblies; SPARK code for 3D core steady-state and transient analysis; and LtoS code for assembly homogenization parameter function, which links LOCUST and SPARK. Bamboo-C has all the necessary analysis functions for the fuel management and nuclear design of PWRs, mainly including the start-up physics tests, calculations of the neutron-kinetics parameters, differential and integral worth of rod cluster control assemblies (RCCAs), and power-operation following simulation. Finally, the engineering validations of Bamboo-C have been completed according to the operation data from the reactors CNP300, CNP650 and CNP1000 designed by China independently. The validation results show that the errors between the values of such key parameters of cores as critical boron concentration, temperature coefficient, RCCA worth, and power distributions, calculated by Bamboo-C, and their measured values satisfy the corresponding engineering criterion limits.

Comparative Analysis of Genetic Algorithms Based on Different Selection Strategies for Refueling Optimization in the Ratio Method

Li Zhan, Zhou Xuhua, Ding Ming, Huang Jie

2021, 42(5): 23-29. doi: 10.13832/j.jnpe.2021.05.0023

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Abstract:
The genetic algorithm is one of the classic algorithms applied to the refueling optimization. An important part of this algorithm is the selection strategies. The existing studies often directly adopt the roulette wheel selection and stochastic tournament selection, and are lacking in comparison and analysis of different selection strategies. To obtain the selection strategy with the strongest optimization capability, this study, with the 1/6 core of a thorium-based prismatic high-temperature gas-cooled reactor (HTGR) taken as an example, constructs the fitness function in the ratio method, performs core physics calculation using the DRAGON code, and in conjunction with the elitism strategy, compares the optimization capabilities of the five selection strategies, including the roulette wheel selection, stochastic tournament selection, uniform ranking method, exponential ranking selection and deterministic selection. The study results show that the optimization capability of the exponential ranking selection is superior to the other four strategies, so the exponential ranking selection is most suitable for solving the refueling optimization problems.

Numerical Simulation of Turbulent Mixing of LBE between Sub-Channels of Wire-Wrapped Fuel Assembly

Wang Jingjie, Zhu Dahuan, Lu Tao, Deng Jian, Cai Rong

2021, 42(5): 30-35. doi: 10.13832/j.jnpe.2021.05.0030

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Abstract:
The lead-bismuth eutectic (LBE) is a candidate coolant for the fourth generation liquid metal reactors. In view of its special thermophysical properties, its flow and heat transfer process in the fuel assembly sub-channels need to be studied. For this purpose, the authors conduct numerical simulation and analysis of the turbulent flow of the LBE in the wire-wrapped fuel assembly, and compare the numerical simulation results of the wall temperature of fuel rods with the experimental data for response. The simulation results agree well with the experimental data, indicating that the mathematical model and numerical results are highly reliable and accurate. The authors also characterize the turbulent mixing of LBE between different sub-channels at different values of the ratio of spacing between fuel rods to the fuel rod diameter (S/D), using the turbulent mixing coefficient β. The characterization results show that the fluctuation of β between different sub-channels varies and that the β value is negatively correlated with the S/D. Based on different calculation results of the S/D and Renolyd number, the correlation formula for the β between different sub-channels is obtained by fit. It provides a turbulent mixing model for the development of a sub-channel analysis code of fuel assembly consisting of wire-wrapped fuel rods arranged in triangular form.

Comparative Analysis of Critical Velocities of Inter-tube Oscillation and Reverse Flow in Inverted U-tube Steam Generator under Single-phase Conditions

Zhang Rui, Ma Zaiyong, Jiang Zhipeng, Zhang Luteng, Tang Yu, Yue Nina, Sun Wan, Pan Liangming, Zhou Wenxiong

2021, 42(5): 36-41. doi: 10.13832/j.jnpe.2021.05.0036

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Abstract:
In order to improve the safety and cost-effectiveness of nuclear reactor system, the authors study experimentally the relationship between the inter-tube oscillation criticality and reverse criticality in inverted U-shaped tube under single phase condition. By processing the experimental data, the authors then obtain and compare the critical velocities of inter-tube oscillation and reverse flow under different conditions. The results show that under the experimental conditions, the critical velocity of inter-tube oscillation is always higher than that of reverse flow, and the ratio of the former to latter one reaches up to 1.46. It increases with the rise of the inlet temperature of the primary side and the fall of the loop resistance, as well as with the increase of the cooling water flow on the secondary side, but at a gradually decreasing rate. Also, the results indicate that the loop resistance can significantly limit the oscillation, and that a serious inter-tube oscillation may occur at a low loop resistance.

Study on High-Fidelity Thermal-Neutronic Coupling Method Based on the Unified Geometry Modeling and its Application in Experimental Reactor Core Calculation for SPERT

Zhang Minwan, Liu Zhouyu, Wang Bo, Cao Lu, Zhao Chen, Cao Liangzhi

2021, 42(5): 42-50. doi: 10.13832/j.jnpe.2021.05.0042

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To solve the problem that the thermal-neutronic grid mapping relation is complicated and cannot be preset in a centralized manner due to the existence of irregular geometry in performing the thermal-neutronic coupled simulation calculation for various small power reactors and experimental reactors, this paper studies the thermal-neutronic coupling method based on the unified geometric modeling, using the high-fidelity numerical code for reactor neutronics calculation, NECP-X. This study establishes the mapping relation for the thermal-neutronic coupling on the basis of the neutronics model, and enables the direct transient calculation of the experimental reactor core for the special power excursion reactor test (SPERT) via combination with the transient calculation method in NECP-X. Then, this study calculates the steady-state case for the experimental reactors of SPERT, and compares the calculation results with the results gained from the Monte Carlo code. On this basis, this study conducts transient calculation and analysis for these experimental reactors and compares the corresponding results with the experimental results. The final results show that the eigenvalues from the neutronics calculation by the NECP-X and the rod power distribution calculation results are of high accuracy; that the grid mapping method based on the unified geometric modeling allows a easy and fast thermal-neutronic coupled calculation of the PWRs of complex geometry; and that compared with the experimental values, the curve of change in the total power and reactivity gained from transient calculation with time is more accurate and can provide refined distributions of power and temperature.

Study on Selection of Internal Initiating Events of Small Natural Circulation Lead-Cooled Fast Reactor

Zhao Pengcheng, Liu Zijing, Li Jie, Yu Tao

2021, 42(5): 51-56. doi: 10.13832/j.jnpe.2021.05.0051

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Abstract:
The initiating events, as the starting point and basis for the deterministic safety analysis and probabilistic safety assessment of the lead-cooled fast reactor (LFR), provide important guidance on the design optimization and safe operation of reactors. This paper, based on the current design scheme of the small natural circulation lead-cooled fast reactor SNCLFR-100 and by reference to the experience of selecting initiating events of other advanced fast reactors, with the generalized “core melting” taken as the top target event, provides the internal initiating events of SNCLFR-100 by deduction in the master logic diagram (MLD) method, which form a relatively complete list of internal initiating events. This study thus can provide a theoretical basis for the safety analysis of natural circulation LFRs.

Study on Heat Transfer Characteristics and Optimization of Spent Fuel Storage and Transportation Containers Based on Heat-Fluid-Solid Coupled Model for Radiation, Convection and Conduction

Long Teng, Zhang Guihe, Jin Gang, Deng Xiaoyun, Kong Xiaofei, Jin Ting, Xiong Guangming

2021, 42(5): 57-63. doi: 10.13832/j.jnpe.2021.05.0057

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Based on the heat-fluid-solid coupled computational fluid dynamics (CFD) model for radiation,convection and conduction, this paper compares the changes in maximum temperature, natural convection flow rate, outer surface radiation and convection power of various parts under different storage status and ambient temperatures, with different packing media in the containers, and with or without annular fin. According to the study results, horizontal arrangement improves the convection heat transfer on the annular fin; each time the ambient temperature increases by 10 ℃, the fuel cladding temperature and outer wall surface temperature increase by 6.5 ℃ and 8.3℃, respectively; when the medium filling the fuel basket is changed from neutron absorption plate or aluminum block to helium, the overall heat transfer capacity of the container deteriorates dramatically, with the temperature inside the container increasing and the outer wall surface temperature remaining almost unchanged; with the solar radiation ignored, a annular fin will help improve convection heat transfer and thus decrease the overall temperature of the container, otherwise, the annular fin absorbs more solar energy, causing the overall temperature of the container to exceed that of a container with smooth wall; and the CFD radiation power is basically same as the calculation results gained in the formula method, as shown in the verification of the CFD radiation model based on the algebraic analysis method and diffuse gray surface model.

Experimental Study on Sensitivity of PRHR Pipeline Break Location on ACME Test Facility

Liu Yusheng, Xu Chao, Wu Peng, Wang Nan, Li Zhenxiao

2021, 42(5): 64-70. doi: 10.13832/j.jnpe.2021.05.0064

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Abstract:
To study the safety performance of the advanced passive (AP) nuclear power plant (NPP) under passive system failure condition, the experimental study on the loss of coolant from the passive residual heat removal (PRHR) pipeline break is performed by the advanced core-cooling mechanism experiment (ACME) facility, during which the effect of main test sequences and break location on the key parameters in different phases of the accident is analyzed. As demonstrated by the study results, the ACME PRHR pipeline break test sequences, basically same as those for the small-break loss of coolant accident (SBLOCA) of the cold leg, reproduce the safety characteristics in the natural circulation phase of the passive NPP, blowout phase of the automatic depressurization system (ADS) and safety injection (SI) phase of the in-containment refueling water storage tank (IRWST); in the test at different break locations, the passive core cooling system (PXS) can always ensure the water makeup for core and the core active region remains below the mixed liquid level; and the break locations have notable effect on the ACME LOCA accident sequence, initial depressurization rate of reactor coolant system (RCS), PRHR heat exchanger (HX) flow, blowout flow, core level, IRWST SI flow and other parameters, and have little impact on the SI flow of the core makeup tank (CMT) and accumulator (ACC).

Study on the Effect of MA Nuclides Transmutation on Safety in Lead-Cooled Fast Reactors

Fu Peng, Liu Bing, Zhang Xinying

2021, 42(5): 71-75. doi: 10.13832/j.jnpe.2021.05.0071

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The lead-cooled fast reactor can be used for the post-processing of part of the minor actinides (MA) nuclides contained in the spent fuel. This study designs three modes of adding MA nuclides to analyze and study the effect of the MA nuclides transmutation on the core critical performance, core life cycle and fuel temperature coefficient thus to study the effect of the MA nuclides addition on the reactor safety performance. The results show that the addition of MA nuclides reduces the initial critical performance of the core; that the addition by either coating or mixing with fuel can significantly extend the life cycle of the lead-cooled fast reactor, while the addition of transmutation rod has different effect on the core life cycle depending on the rod location; and that the addition of MA nuclides causes the change of fuel temperature coefficient, which, however, remains negative. All of the three addition modes are feasible. In particular, attention should be paid to the effect of the transmutation rod location on the core life cycle. It is not advisable to distribute the transmutation rods in a concentrated area.

Analysis of Post-Accident Debris Transport Performance in In-Containment Refueling Water Storage Tank in Nuclear Power Plants

Hou Jianfei, Wang Qingli, Si Hengyuan

2021, 42(5): 76-80. doi: 10.13832/j.jnpe.2021.05.0076

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Abstract:
To avoid the blockage of strainer in the in-containment refueling water storage tank (IRWST) and guarantee the safe operation of the pump downstream the IRWST after accident occurrence, the debris transport performance in the IRWST must be evaluated carefully. For the double-loop pool-type IRWST of a nuclear power plant (NPP), the Computational Fluid Dynamics (CFD) method is adopted to simulate the flow field of this IRWST, and the volume ratio of the high velocity region and high turbulent kinetic region is used to quantitatively evaluate the debris transport performance of the IRWST. The results indicate that the debris transport ratio of IRWST does not exceed the design value of the strainer under various post-accident conditions, which ensure the safety of the strainer and its connected systems after accident occurrence; the strainer load reaches the maximum value when only the strainer A in the inner loop is put into operation; and the post-accident debris transport performance is mainly affected by the high velocity region in the flow field. As a result, this paper proposes the optimization scheme, namely, increasing the diameter of the mixing pipeline in the outer loop, for the existing IRWST layout. This scheme can reduce greatly the debris transport ratio in the IRWST after accident occurrence and thus improve the post-accident safety of the NPP.

Study on Uncertainty Analysis Method of Fast Reactor Based on Covariance Matrix Sampling

Zhu Runze, Ma Xubo, Wang Dongyong, Zhang Bin, Peng Xingjie, Wang Lianjie

2021, 42(5): 81-85. doi: 10.13832/j.jnpe.2021.05.0081

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Abstract:
The uncertainty analysis methods based on traditional statistical sampling have received widespread attention in China and other countries due to their simple algorithms, easy realization of codes, and consideration of high-order effects. However, these methods usually require a large number of samples to ensure the calculation accuracy of response variables. As found in the study, this phenomenon occurs because of the poor quality of the samples. After a covariance matrix sampling is used instead of the traditional sampling method, a small sample size can also ensure a high calculation accuracy. This paper firstly demonstrates theoretically the feasibility of the covariance matrix sampling method, and verifies it with simple tests. On this basis, this paper, using the self-developed fast spectrum reactor sensitivity and uncertainty analysis code - SUFR and the international reference configuration for fast reactor ZPR-6/7, calculates the uncertainty of effective multiplication factor (k_eff) caused by the nuclear cross sections of different reaction types of multiple nuclides, and compares the calculation results with the uncertainty calculated using the deterministic method. As demonstrated by the results, if the covariance matrix sampling is used, with a sample size of 50, the uncertainty deviation calculated in the two methods each is below 1.3%. This indicates that the use of the covariance matrix sampling method can solve the problems present in the use of the traditional sampling method to calculate uncertainty, and that it is appropriate to develop the SUFR code function against the covariance matrix sampling. This method represents a further development of the traditional sampling method.

Study on Implementation Method of Deep Subcritical Rod Worth Measurement Electronics

Luo Tingfang, Zhu Hongliang, Gao Zhiyu, Bao Chao, Wang Yinli, Qing Xianguo, He Zhengxi, Sun Qi, Yang Zhenlei, Yuan Hang, Shan Wei

2021, 42(5): 86-89. doi: 10.13832/j.jnpe.2021.05.0086

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In order to realize the effective acquisition of data required for deep subcritical rod worth calculation, the overall architecture and key modules of deep subcritical rod worth measurement electronics have been researched and designed. The characteristics of the key modules have been tested through reactor test. The results show that the designed deep subcritical rod worth measurement electronics can effectively measure the detector signal transmitted through about 200 m cable. The pulse signal has a stable waveform width and a proper signal-to-noise ratio. The measured high voltage plateau characteristic curve can provide an effective reference for the high voltage selection of the detector. The measured discrimination characteristic curve is stable, from which the neutron component in the detector signal can be effectively obtained.

Experiemental Study on Heat Transfer of Supercritical Water in Triangular Channel of Reactor Core with Spacer Grid

Wang Weishu, Huang Zhihao, Xu Weihui, Ma Ziqiang, Zhu Xiaojing, Bi Qincheng

2021, 42(5): 90-95. doi: 10.13832/j.jnpe.2021.05.0090

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Abstract:
The experimental study on supercritical water flow heat transfer is carried out for the vertical upward triangular sub-channels in the supercritical water cooled reactor core with spacer grid. The fuel bundle diameter and pitch ratio of this sub-channel used are 8 mm and 1.4 respectively. The parameters adopted in the research include heat flux (q) ranging from 600 kW/m², pressure (P) ranging from 23-28 MPa, and mass flow rate (G) ranging from 700-1,300 kg/(m²·s). This study analyzes the effect of such thermal parameters as heat flux, pressure and mass flow rate on the heat transfer characteristics of supercritical water. According to the experimental results, with the mass flow rate at the spacer grid increasing, the fluid disturbance increases and the heat transfer coefficient rises significantly; under the supercritical pressure, the inner surface temperature increases with the increasing pressure, and consequently, the peak heat transfer coefficient decreases; as the peak heat transfer coefficient reduces at an excessively high heat flux, the heat transfer performance can be improved when the heat flux is reduced properly; increasing the mass flow rate causes the drop of inner surface temperature and the increase of peak heat transfer coefficient, and thus can significantly improve the heat transfer performance; and the pressure change has little effect on the heat transfer characteristics in the vicinity of spacer grid, and however, the system safety can be improved by increasing the pressure properly.

Experimental Study on Flow Mixing Characteristics of Lower Plenum of Small PWR

Wang Chunyu, Peng Fan, Xing Jun, Wang Long, Xiao Weiming

2021, 42(5): 96-102. doi: 10.13832/j.jnpe.2021.05.0096

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In order to study the flow mixing characteristics of lower plenum of small pressurized-water reactor (PWR), this study conducts a hydraulic experiment on the reactor model in scale of 1∶3 based on the scaling method, and obtains the flow mixing factor matrices for the core inlet under the balanced and unbalanced flow conditions of the cold leg by measuring the solution concentration change. The results show that under the balanced flow condition, the changes of cold leg flow have slight effect on the core inlet flow mixing factor matrix, while under the unbalanced flow condition, the fuel assembly flow mixing factor near the outlet pipe is significantly affected by the flow unbalance, and that small change occurs to the flow mixing factor in the central region. Therefore, the flow mixing characteristics of lower plenum of small PWR are stable at the balanced flow, and focus should be put on the change in flow mixing characteristics of fuel assemblies near the outlet under the unbalanced flow condition.

Study on Activation-Induced Dose Calculation of Reactor Structural Materials

Yuan Xudong, Ma Huiqiang, Chen Zhenping, Guo Shuwei, Xie Jinsen, Yang Chao, Yu Tao

2021, 42(5): 103-109. doi: 10.13832/j.jnpe.2021.05.0103

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Numerous radioactive nuclides are produced from the neutron activation reaction of reactor structural materials under the core neutron irradiation. The decay photon from activated reactor structural materials is an important source of occupational exposure dose for workers during reactor shutdown & maintenance, refueling or decommissioning. This paper studies the cell activation calculation method for reactor structural materials based on Rigorous two step (R2S) method, and develops a reactor structural material activation dose calculation program, MOCA, based on the Monte Carlo particle transport code MCNP and point activation calculation code ORIGEN. Then, this study develops the functional interface and data interface codes to realize the automatic coupling of the transport code and activation calculation code thus to realize fully automated “neutron transport-activation analysis - dose calculation” coupled analysis. Finally, the M5 fuel cladding activation calculation model, stainless steel activation calculation model and NUREG/CR-6115 PWR model are used to perform benchmark verification of the MOCA, which proves that the MOCA is adequate and reliable.

Irradiation Test and Performance Evaluation of N36 Characteristic Fuel

Zhang Kun, Chen Ping, Xing Shuo, Pang Hua, Peng Hang, Pu Zengping, He Liang, Zhang Lin, Qiu Bowen

2021, 42(5): 110-113. doi: 10.13832/j.jnpe.2021.05.0110

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N36, a kind of advanced zirconium alloy developed independently in China, will be adopted as the cladding material of fuel assembly in HPR1000. In order to study the N36 alloy cladding performance in a reactor and verify the feasibility of using this alloy for cladding, the N36 characteristic fuel assembly with the N36 alloy cladding is designed and loaded in the reactor of Qinshan NPP Phase II for irradiation test. The pool side examination is performed at the end of each cycle to collect the in-core performance data of this fuel assembly. Then, the performance of N36 alloy cladding is analyzed and evaluated based on these data. For this sake, this study provides the irradiation test scheme, design and in-core performance collection method of the N36 characteristic fuel, and compares the in-core performance of N36 alloy and Zr-4 alloy.

Study on Crack Impact Toughness Evaluation Method for Metallic Materials

Li Yilei, Li Pengzhou, Yao Di, Qiao Hongwei, Zhang Kun, Sun Lei

2021, 42(5): 114-118. doi: 10.13832/j.jnpe.2021.05.0114

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For the engineering structures in service for a long period in the complex environment, crack initiation and growth are difficult to avoid. Therefore, for the engineering structures required to withstand explosion and impact, the crack impact toughness of their materials must be evaluated to prevent the possible brittle fracture under impact due to crack occurrence. This study, relying on the Instron VHS high strain rate material testing machine, develops a set of dynamic fracture testing device to measure the crack ductile-brittle transition process of 4 different kinds of metallic material with high impact energy under impact, and studies the factors influencing the crack ductile-brittle transition rate of these materials. According to the findings, the Charpy impact energy can not fully reflect the crack impact toughness,and the absence and existence of preformed cracks, the specimen constraint mode and the specimen crack tip constraint factor all have an effect on the ductile-brittle transition rate in the crack impact test of metallic materials. Based on the results above, this study finally propose the basic ideas about the crack impact toughness evaluation method for metallic materials.

Study on Heat Distribution Characteristics of Direct Safety Injection of Reactor Pressure Vessel

Jiang Xing, Weng Yu, Wang Haijun

2021, 42(5): 119-122. doi: 10.13832/j.jnpe.2021.05.0119

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In the passive PWRs in China, the cooling water injection pipe for emergency cooling system is directly connected to the pressure vessel. Unlike the traditional safety injection (SI) of cold leg, this SI mode is called reactor direct SI. For the reactor pressure vessel under the SI conditions, this paper studies the heat distribution shape of the SI fluid on the surface of the pressure vessel by the combination of physical experiment and numerical analysis. As shown in the study, unlike the traditional oblique nozzle SI mode for cold leg of the main pipe, the distribution of SI fluid in the downcomer annulus is approximately in the shape of an isosceles triangle under the direct SI condition. Based on the experimental results and the numerical calculation and verification, the heat distribution angle of the pressure vessel is found directly proportional to the flow rate ratio, and moreover, the calculation model of the SI fluid distribution is proposed for reference in the reactor safety design.

Study on Room Temperature Fracture Behavior of Main Pipeline Materials at Medium and Low Loading Rate

Li Pengzhou, Li Yilei, Yao Di, Sun Lei, Qiao Hongwei

2021, 42(5): 123-127. doi: 10.13832/j.jnpe.2021.05.0123

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Cracks may occur at the base metal and welding joint of the main pipeline of a nuclear power plant (NPP) during the long-term service. Therefore, it is necessary to study the fracture behavior of the main pipeline material and welding material at medium and low loading rate to avoid the possible double-ended guillotine break of the main pipeline under strong earthquake impact. This study, relying on the Instron VHS high strain rate testing machine for materials, develops a set of testing method for fracture behavior of materials at medium and low loading rate, which in turn is used to measure the room temperature fracture behaviors of main pipeline nitrogen-containing material (00Cr17Ni12Mo2) and welding material (OK Tigrod 316L) within the loading rate of 0.5 m/s in a NPP. As observed from the results, under the room temperature, the main pipeline nitrogen-containing material (00Cr17Ni12Mo2) shows no crack initiation within the impact rate of 0.5 m/s, and the fracture toughness of the welding material (OK Tigrod 316L) indicates no obvious regular change within the loading rate of 0.5 m/s.

Generation Method of Target Power Spectral Density for Seismic Design of Nuclear Power Plant Equipment

Xie Haoyu, Zhu Yizhou, Zhang Wengang, Tang Guangwu, Xie Yongcheng

2021, 42(5): 128-133. doi: 10.13832/j.jnpe.2021.05.0128

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The input ground motion for the seismic design of nuclear power plant (NPP) equipment is required to envelop both the required response spectrum (RRS) and target power spectral density (PSD). However, no unified algorithm exists to generate the target PSD, whether at home or abroad. Based on the generation methodology for target PSD suggested in Section 3.7.1 of the Standard Review Plan (SRP) of U.S. Nuclear Regulatory Commission, this paper proposes an improved method for target PSD synthesis with the original iteration process optimized. The proposed method is applied successfully to the two RRS cases for nuclear equipment. The result shows that the improved method for target PSD generation is highly compatible with RRS and provides simple and rapid calculation, while its convergence accuracy is similar to the calculations by random vibration theory (RVT). This method is considered as the target PSD test basis for the input artificial ground motion (AGM) for the seismic design of NPP equipment.

Study on Shell Mode Vibration Characteristics of Reactor Core Barrel Based on Analysis of Pressure Vessel Vibration Signal

Luo Ting, Yang Taibo, Liu Caixue, Luo Neng, Hu Jianrong, Jian Jie, Feng Jintao

2021, 42(5): 134-137. doi: 10.13832/j.jnpe.2021.05.0134

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The vibration state of reactor core barrel is directly related to the safe operation of the core. However, the barrel is under high temperature and high radiation conditions, so it is impossible to measure the barrel vibration by a sensor directly installed on the barrel. In this paper, the accelerometer installed on the pressure vessel is used to indirectly monitor the barrel vibration. The coherence spectrum, auto-power spectrum and cross-power spectrum of vibration signal of pressure vessel with multiple nuclear units are analyzed to provide the shell mode vibration frequency and amplitude of the barrel. As shown by comparison, the analysis results approximate to the measured values from the test of Qinshan NPP Phase II Unit 1. The study results show that the shell mode vibration characteristics of the core barrel can be effectively identified by the monitoring and analysis of the pressure vessel vibration signals, which provides a foundation for the evaluation of the core barrel conditions.

Cause Analysis and Solution of Main Steam Pipe Vibration in a Nuclear Power Plant

Xia Shuan, Zhan Minming, Chen Xingwen, Zhan Kai, Wu Xinzhuang

2021, 42(5): 138-142. doi: 10.13832/j.jnpe.2021.05.0138

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During the power increase test and full power operation, a nuclear power plant (NPP) found that the noise in the main control room (MCR) exceeded the design target value. After measurement, the vibration of the main steam pipe and its transmission to the MCR through the supports and penetrations are considered as one of the main reasons for the excessive noise in the MCR. In view of this situation, this paper uses the fluid dynamics code and acoustic analysis code to analyze the flow field and sound field of the main steam pipe in the flow-sound coupled analysis method. Acoustic resonance is found at the branch pipe of the main steam safety valve and in the cavity of the main steam isolation valve (MSIV), which is the main cause for the main steam pipe vibration. Considering the causes for the main steam pipe vibration, the MCR noise can be reduced by vibration source control, transmission path control and other means.

Study on Multivariable Decoupling Control of Small PWR Pressurizer based on Active Disturbance Rejection Control

Shi Bo, Li Daixing, Guo Wei, Zhang Yilin

2021, 42(5): 143-148. doi: 10.13832/j.jnpe.2021.05.0143

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Abstract:
Since the pressurizer (PZR) in a small pressurized water reactor (PWR) features non-linearity, time-dependence, and strong coupling, its accurate mathematical model is hard to build and the traditional control method cannot provide a satisfactory control effect. Therefore, this study describes a multivariable decoupling control method based on the active disturbance rejection control (ADRC) technique for such PZR. In this study, a non-equilibrium three-region model of PZR is built, and processed linearly based on the perturbation theory to generate the pressure and water level coupled transfer function equation. Then, an ADRC based decoupling controller is designed based on the transfer function, and the parameters of the controller are subjected to multi-objective optimization via differential evolution. Finally, the ADRC based decoupling control and traditional proportional-integral-derivative (PID) control for the same small reactor PZR are compared and analyzed by the MATLAB simulation platform. As demonstrate by the comparison, the ADRC based decoupling controller can effectively solve the PZR pressure and level coupling problem and has better robustness and interference immunity than the traditional PID controller. Therefore, this study provides a theoretical foundation for the engineering application of the ADRC method to the PZRs.

Study on Data Driven Anomaly Detection and Analysis Algorithm for Nuclear Power Systems

Wang Xiaolong, Zhang Yongfa, Liu Zhong, Cai Qi, Zhao Xin, Zheng Jintao

2021, 42(5): 149-155. doi: 10.13832/j.jnpe.2021.05.0149

Abstract(295) HTML (119) PDF(71)

Abstract:
Aiming at the problem that there are few and incomplete fault samples in nuclear power system on-line anomaly detection, referring to the concept of Safe operation domain, and based on the idea of logical distance calculation, this paper proposes a nuclear power system anomaly detection algorithm based on normal operation data. Taking the historical data of common operating conditions of a nuclear power system as the object, numerical experiments were carried out to verify the algorithm. The results show that the design algorithm can effectively detect system anomalies and faults, with good reliability and interpretability, and the detection strength can be adjusted.

Assessment of No-Core-Melt Concept for Pressure Tube Supercritical Water Cooled Reactors under Extreme Accidents

Wu Pan, Ren Yanhao, Shan Jianqiang, Huang Yanping

2021, 42(5): 156-161. doi: 10.13832/j.jnpe.2021.05.0156

Abstract(267) HTML (79) PDF(23)

Abstract:
This paper develops and verifies the two-dimensional (2D) heat conduction model and radiation heat transfer model based on the self-developed accident analysis code SCTRAN, applies the improved SCTRAN code to the core safety assessment of the Canadian pressure tube supercritical water cooled reactor (PT-SCWR) under the loss-of-coolant accident (LOCA) plus loss of emergency core cooling system (LOECC) accident, and assesses the heat transfer efficiency between the fuel rods and moderator and the key factors. The assessment results show that the residual decay heat of the reactor can be effectively removed by the radiation heat transfer from the fuel rods to fuel channels and the natural convection heat exchange from the fuel rods to steam and that the maximum cladding temperatures of the fuel rods in inner and outer rings of the fuel assembly at the maximum power are 1,278℃ and 1,192℃, respectively, which are below the stainless steel cladding melting temperature. Therefore, no core meltdown occurs throughout the accident.

Study on Function Extension of Pressurizer Fast Depressurization System for Gen Ⅲ PWR NPPs

Zhang Jiajia, Deng Wei, Xiao Jun, Gong Yu

2021, 42(5): 162-166. doi: 10.13832/j.jnpe.2021.05.0162

Abstract(306) HTML (187) PDF(32)

Abstract:
A nuclear power plant of Gen Ⅲ pressurized water reactor (PWR) in China has installed the pressurizer fast depressurization system for fast depressurization of the primary loop under severe accidents. With respect to this NPP, according to the general method and process of applying the probabilistic safety assessment (PSA) to the NPP design improvement, this study, focusing on the improvement plan of extending the pressurizer fast depressurization system function to the primary feed & bleed for depressurization to enable the use of this system as a standby pressure relief measure for the pressurizer safety valve, performs the PSA modeling and the feasibility evaluation and demonstration. Finally, this study shows that the improvement plan can reduce the core damage frequency of a NPP greatly, without any new negative effect, so the plan is feasible and can be implemented. Consequently, the NPPs should fully tap the potential of existing system equipment to further improve the safety and economical efficiency of NPPs.

Analysis of Mal-Operation Accidents of Nuclear Power Plant I&C System

Cai Wei, Bao Guogang, Yue Zhidong, Lu Changdong

2021, 42(5): 167-172. doi: 10.13832/j.jnpe.2021.05.0167

Abstract(296) HTML (81) PDF(28)

Abstract:
A method based on the simplified analysis was proposed in order to comprehensively evaluate the accidents of spurious actuation of Instrumentation and Control (I&C) systems of the Nuclear Power Plant (NPP). Based on the concept of “functional group”, the Postulated Initiating Events (PIEs) of spurious I&C actuation were systematically identified and grouped to obtain the potential additional accidents that cannot be bounded by the existing accident analysis. Then these potential additional accidents were qualitatively assessed and quantitatively analyzed according to the conservative analysis assumptions and rules. The results show that the protection systems of the NPP can provide diverse protection against the spurious I&C accidents and the consequences meet the acceptance criteria. Besides, the function of “startup of safety injection and trip of the main pumps triggered by the low 2 signal of the difference between the local pressure and saturation pressure in two hot legs” was suggested to be added.

Reliability Modeling and Analysis of Reactor Protection System Based on FPGA

Zhang Dongliang, Zhang Kaiwen, Zhang Chaofan

2021, 42(5): 173-177. doi: 10.13832/j.jnpe.2021.05.0173

Abstract(490) HTML (215) PDF(48)

Abstract:
In order to establish a reactor protection system reliability model based on the field programmable gate array (FPGA) to provide effective analysis and verification methods for system safety, this study adopts the fault tree and stochastic Petri net (SPN) models to perform reliability modeling and analysis for the single channel of CANDU reactor shutdown system 1 (SDS1). The analysis by fault tree model provides the minimal cut set. With the top event probability taken as the system fault probability, and in consideration of the fault detection, maintenance and periodic testing, the probability of rejection of the system is obtained by the stochastic Petri net model simulation. The results show that both the fault tree and state space representation are limited to a certain extent, while the stochastic Petri net can reflect the impact of fault detection and periodic tests on the reactor protection system, dynamically present the system reliability, and avoid the problem of state space explosion. Therefore, the stochastic Petri net model established in this study is suitable for the reliability modeling of the reactor protection system.

Study on Gas Relaxation Process in Leakage Rate Measurement Stage of Containment Test Based on Time Series Analysis Method

Shen Dongming, He Rui, Zhang Ji

2021, 42(5): 178-181. doi: 10.13832/j.jnpe.2021.05.0178

Abstract(290) HTML (131) PDF(28)

Abstract:
During the containment test, upon the leakage rate measurement plateau, the temperature, steam partial pressure, and other parameters of gases in the containment fluctuate and then go into the steady relaxation process. This article analyzes the criteria for reaching new steady state, the gas relaxation time, the factors that affect the relaxation process and other aspects to provide the criterion for steady state by performing steady state test via test statistic calculation. Also, this article conducts time series decomposition for various parameters involved in the relaxation process to analyze the relaxation process of each parameter separately. As seen from the study results, the relaxation time is mainly affected by the non-uniform steam partial pressure due to containment pressurization. On this basis, this study further proposes controlling the steam partial pressure imbalance trend to shorten the relaxation time of the leakage rate results.

Analysis of Sentitivity of Fission Product Iodine in Containment to Various Factors under Severe Accidents

Hu Wenchao, Pan Xinyi, Zhang Pan, Zhao Chuanqi, Sun Haixu, Yi Yan

2021, 42(5): 182-188. doi: 10.13832/j.jnpe.2021.05.0182

Abstract(174) HTML (136) PDF(48)

Abstract:
The most serious consequence of reactor accident is that the radioactive fission products are dispersed into the environment. To study the distribution characteristics of radioactive fission product iodine in the containment under severe accident conditions, this paper assumes that the fission product iodine is released from the primary system into the containment due to the occurrence of severe accident in the nuclear power plant (NPP). Then, using the accident source term evaluation code (ASTEC), this study builds a containment construction model for the NPP, sets the boundary conditions, and calculates the chemical form, chemical properties, distribution, and change trends of different compounds, of the fission product iodine under different pH values, with or without silver (Ag) injection and under gaseous radiation conditions. The results show that the production of volatile iodine in the containment can be inhibited under alkaline conditions; the silver can promote the iodine trap in the liquid phase and reduce the iodine volatility; and the gaseous radiation environment can promote the formation of gaseous CH₃I and IO_x. As a result, this study can provide guidance for the removal of radioactive iodine in the containment under serious accidents.

Numerical Study on the Performance of Oxygen Control Bypass of Lead-Bismuth Eutectic System

Li Xiaobo, Wang Yifeng, Zhu Huiping, Liu Yang, Niu Fenglei

2021, 42(5): 189-194. doi: 10.13832/j.jnpe.2021.05.0189

Abstract(275) HTML (71) PDF(29)

Abstract:
In order to explore the method for efficiently regulating the oxygen concentration of lead-bismuth eutectic (LBE) system by solid-phase oxygen control, this study calculates the oxygen concentration of LBE system by the self-programmed code in FORTRAN language based on the lumped parameter method, by correcting the empirical formula for liquid LBE corrosion and in combination with the PbO dissolution model, and then studies thereby the effects of the main loop flow rate, temperature in the mass exchanger (MX), and PbO inventory on the MX oxygen supply performance of the independently designed small LBE system, to establish the preliminary design criteria of MX thus to obtain the oxygen supply performance of MX and design parameters of oxygen supply bypass under certain constraints. This study can provide a reference for the design and calculation of the oxygen control bypass of LBE system, and also present an efficient approach to calculating the transient oxygen concentration and corrosion of the LBE system to give a new insight into establishing the oxygen concentration model prediction system.

Tracking and Identifying CRDM Degradation State Based on SWT and Phase Space Warping

Zhu Kang, Zhao Xinwen, Zhang Liming, Yu Hang

2021, 42(5): 195-199. doi: 10.13832/j.jnpe.2021.05.0195

Abstract(199) HTML (87) PDF(26)

Abstract:
The control rod drive mechanism (CRDM) is important for ensuring the normal operation and safety of reactors. Considering the low speed, latent performance degradation signs, and susceptibility to interference from other signal sources of the CRDM, this paper tracks and identifies the CRDM degradation state by a combination of the semi-soft wavelet threshold (SWT) and phase space warping (PSW). By comparison with the method based on the root mean square (RMS) of vibration amplitude, the method combining the SWT with PSW, unlike the traditional linear signal processing theory with poor performance in tracking and identifying the latent performance degradation state, can track and identify adequately the performance degradation state of rotating machinery with high background noise.

Study on Thermal Stratification of Marine Reactor Surge Line Based on Operating Data

Jiao Meng, Cai Qi, Zhang Yongfa, Wang Xiaolong, Jiang Lizhi

2021, 42(5): 200-205. doi: 10.13832/j.jnpe.2021.05.0200

Abstract(203) HTML (106) PDF(37)

Abstract:
The thermal stratification of marine reactor surge line is divided into four types of typical transient based on the operating data: power increase, power decrease, off-design conditions, and small spray flow. The dimensionless Richardson number (Ri) analysis and the numerical simulation calculation of transient conditions are performed separately for these typical transients, generating the thermal stratification section length, duration and maximum temperature difference along the horizontal line section of the surge line under these typical transients. As indicated by the results, the thermal stratification under power increase and decrease transient conditions extends through the surge line just once, and the surge line safety may be affected by the circulating thermal fluctuation at the joint under the power increase transient condition, the thermal stratification of long section, long time and large temperature difference in the horizontal section under the small spray flow transient condition, and the thermal stress fluctuation caused by the off-design conditions. The method of study on thermal stratification of surge line based on operating data, proposed in this paper, lays a foundation for the subsequent thermal stress and thermal fatigue analyses, and also provides reference for the thermal stratification studies of other volumetric equipment.

Study on Electromagnetic Structure Design of Electromagnetic Bearing Body for Control Rod Drive Mechanism

Yu Tianda, Peng Hang, Wu Hao, Du Hua, Li Wei, Tang Jiankai, Song Liwei

2021, 42(5): 206-212. doi: 10.13832/j.jnpe.2021.05.0206

Abstract(201) HTML (36) PDF(24)

Abstract:
The electromagnetic structure design is the core of the design of electromagnetic bearing body for control rod drive mechanism (CRDM). In order to obtain a reasonable electromagnetic structure, a certain electromagnetic structure is selected according to different shapes of magnetic flux distribution, and used to build the analytical model and finite element model for electromagnetic analysis. Then, the working point of the electromagnetic bearing is determined by finite element simulation and analysis, and the key performance parameters, such as current stiffness and displacement stiffness, of the electromagnetic bearing are further analyzed and studied. Finally, cross verification is performed by comparison between the finite element simulation and analysis results and the analytical calculation results. The verification results show that the electromagnetic structure design of the CRDM electromagnetic bearing body is appropriate, and all performance indexes can meet the design requirements.

Study on Electromagnetic Structure Design of Linear Motor Type Reactor Control Rod Drive Mechanism

Yu Tianda, Peng Hang, Deng Qiang, Li Wei, Wu Hao, Fu Guozhong, Tang Jiankai

2021, 42(5): 213-217. doi: 10.13832/j.jnpe.2021.05.0213

Abstract(188) HTML (48) PDF(37)

Abstract:
The electromagnetic structure design is the core of the design of control rod drive mechanism (CRDM) for electromagnetic reactors. This study introduces a cylindrical linear motor suitable for the CRDM. Firstly, this study describes the basic electromagnetic structure of such motor. Secondly, in an analytical calculation method, the feasible slot-pole combination is derived and a new concentrated winding structure is analyzed. Thirdly, further study is performed on the key factors affecting the performance indexes of the CRDM, including the slot-pole combination, winding structure form, and electromagnetic structure dimensions, in the 2D finite element method (FEM). Finally, based on the further study results, the detailed design and test & verification of the electromagnetic structure are carried out. The test results show that the linear motor type CRDM is designed reasonably, and can perform the basic functions, such as lifting, inserting downward, holding and dropping, in accordance with the control commands, and that its technical specifications, such as the lifting force, following characteristic and rod drop performance, can meet the design requirements.

Analysis of Reactor Coolant Pump Thrust Bearing Modification of an HPR1000 Unit

Du Pengcheng, Fei Dongdong, Wen Xue

2021, 42(5): 218-221. doi: 10.13832/j.jnpe.2021.05.0218

Abstract(330) HTML (218) PDF(69)

Abstract:
The abnormal thrust bearing abrasion was found during the small flow test of the reactor coolant pump (RCP) of three-bearing structure design for an HPR1000 unit. After analyzing the RCP thrust bearing structure, the authors identify and analyze all the possible causes for this abrasion in the root cause analysis method based on fishbone diagram. According to the identification and analysis results, the authors proposing such modifications as using the oil supply of multi-nozzle design, adding oil suction chamfer at the oil film suction ports of the main and auxiliary thrust bearings, adding jacking oil structure design for reverse auxiliary thrust bearing to the original jacking oil design, build an integrated measurement system for thrust bearing temperature-oil temperature, and supporting the structure by spring plate actively compensated thrust bearing. As verified by tests, the improved RCP thrust bearing system helps enhance significantly the RCP operation reliability and inherent safety of the HPR1000 unit.

Mechanical Reliability Analysis of Control Rod Assembly for HTR-PM

Gao Jianyong, Qing Chen

2021, 42(5): 222-225. doi: 10.13832/j.jnpe.2021.05.0222

Abstract(208) HTML (86) PDF(29)

Abstract:
The article analyzes the forces and weaknesses of the control rod assembly in the high temperature gas cooled reactor (HTGR) demonstration project-pebble bed modular HTGR plant (HTR-PM), using the the mechanical reliability theory and mechanical simulation, and identifies the weaknesses at the connection of the control rod assembly. In the reliability-life calculation method based on the probabilistic fracture mechanics, in combination with the influence of irradiation on the fatigue life model and in consideration of the stress, the authors establish the reliability-life model to calculate the reliability-life of HTR-PM control rod assembly at the irradiation attenuation coefficient of 0.7. The calculation results show that at the reliability level of 0.99, the life of HTR-PM control rod assembly is about 150,000 action cycles, and that the irradiation has a great influence on the life of the control rod assembly. In view of what mentioned above, this paper can provide reference for the equipment reliability management of the HTR-PM control rod assembly.

Study on Chemical Kinetics of Carbon Migration in Superalloys in the Non-Pure Helium Environment in High Temperature Gas Cooled Reactors

Zheng Wei, Li Haoxiang, Yin Huaqiang, He Xuedong, Wang Qiuhao, Ma Tao

2021, 42(5): 226-231. doi: 10.13832/j.jnpe.2021.05.0226

Abstract(307) HTML (174) PDF(48)

Abstract:
The primary coolant in the high temperature gas cooled reactor (HTGR) contains impurities of low content, which will cause serious corrosion of the superalloys in the HTGR operating at the ultra-high temperature. In particular, the carbon migration between the superalloys and non-pure helium has a great influence on the material performance. This study explores the chemical kinetics principle of the carbon migration in the non-pure helium environment, from which the theoretical criteria for material decarburization and carburization are gained. Also, according to the chemical thermodynamics and kinetics principles, this study calculates the oxygen partial pressure and carbon activity in the non-pure helium environment, and indicates that a high partial pressure ratio of CH₄ to H₂O may lead to serious carburization of the alloys. On this basis, this study presents a widely used carbon migration model, “chromium stable phase diagram”, analyzes the chromium activity calculation method, and summarizes the recommended values. Finally, the corrosion behavior of the 10 MW HTGR (HTR-10) designed by Tsinghua University under the actual operating conditions is obtained by calculation based on the chromium stable phase diagram.

Study on Early Warning of Small Leakage in Primary Loop Based on Data Mining

Bai Xiuchun, Qian Hong

2021, 42(5): 232-239. doi: 10.13832/j.jnpe.2021.05.0232

Abstract(165) HTML (98) PDF(24)

Abstract:
An excessive small leakage in the primary loop may cause serious accidents. In order to prevent such accidents, this study proposes an fault early-warning method of the improved Gaussian mixture model (GMM)-grey relational analysis (GRA)-entropy weighting method (EWM) based on the integration of multiple characteristic parameters. In this method, firstly, the mechanism of the dynamic operating characteristics of the small leakage in the primary loop is analyzed, and the early-warning characteristic parameters are thus determined. Then, based on these characteristic parameters, in combination with the EWM and GRA, a multi-parameter integrated early-warning model is established. Finally, the relational analysis and improved GMM algorithm are used to enable effective learning of the statistical features of massive data so that the early-warning thresholds can be self-adapted to different conditions. As this study shows, effective early warning can be realized in this method under variable operating conditions. Compared with single parameter and fixed threshold warning, this method is more stable and provides more accurate, effective and timely warning, and thus, it can provide a reference for the condition monitoring of the primary system.

Design and Study of Backup Cooling Mode for Spent Fuel Pool of CPR1000 Unit

Zhɑnɡ Guohui, Song Hehang, Luo Zhiping

2021, 42(5): 240-244. doi: 10.13832/j.jnpe.2021.05.0240

Abstract(230) HTML (189) PDF(34)

Abstract:
In view of the risk of loss of component cooling water (RRI) facing the reactor cavity and spent fuel pit cooling and treatment system (PTR) of the CPR1000 units in China in certain cases, this paper analyzes the problems of the unit PTR from the perspective of single cooling water source, and considering the existing PTR equipment, creatively designs the backup cooling mode using other cooling water sources. As demonstrated by the analysis and study, this design improves the reliability of continuously cooling the spent fuel pool, and the diversity and redundancy of the PTR cooling modes.

Experimental Research on Effect of Aerosol in Liquid Pool on Size of Bubbles from Submerged Orifice

Zhang Jibin, Lyu Huanwen

2021, 42(5): 245-249. doi: 10.13832/j.jnpe.2021.05.0245

Abstract(258) HTML (45) PDF(21)

Abstract:
The bubble formation from a submerged orifice in a liquid pool is an important phenomenon in the filtration and discharge process of gas in the containment. During filtration, the size of bubbles from the submerged orifice directly affects the rising speed and the gas-liquid contact area of the bubbles, and thus is one of the important parameters affecting the filter filtration efficiency. As the filtration proceeds, the aerosol content in the liquid pool, among others, may affect the bubble size. This paper conducts a visual experiment to study the bubble formation from a submerged orifice in a liquid pool containing BaSO₄ or TiO₂ aerosol and to observe and analyze the change law of the size of bubble from the submerged orifice, thus to gain the mechanism of the effect of aerosol on size of bubble from the submerged orifice. The results show that high temperature liquid pool and TiO₂ can increase the bubble size, but adding BaSO₄ has no obvious effect. Also, the experiment shows the presence of “small air cavity” on the top of the bubbles formed, and the changes of surface tension and “small air cavity” form the main change mechanism of bubble size.

Study on Laser Decontamination Technology for Metal Scraps with Radioactively Contaminated Surfaces

Zhao Wan, Cao Junjie, Wang Shuai, Wen Xiaojun, Zhang Yongling, Ding Ran, Peng Jing

2021, 42(5): 250-255. doi: 10.13832/j.jnpe.2021.05.0250

Abstract(1484) HTML (150) PDF(108)

Abstract:
In order to apply the laser decontamination technology to the clearance or recycling of metal scraps with radioactively contaminated surfaces, this study builds a laser decontamination experimental setup based on the 350 W nanosecond pulsed fiber laser. Then, this study conducts a series of experiments for the key parameters of the laser ablation and decontamination technology, including laser power, pulse duration, frequency, line spacing and scanning speed. According to these experimental results, the rule and the optimum parameters corresponding to different decontamination depths for the laser decontamination technology are obtained. This study in turn conducts verification test of this technology by applying it to the storage rack baseplate of the control rod pool for a nuclear power plant (NPP). The test results show that after the laser decontamination technology is used and the decontamination depth reaches 10 μm, the contamination level of the surface radioactively contaminated by β rays of the sample is reduced to less than 0.8 Bq/cm², and can meet the clearance level.

Design and Study of Critical Physical Test Scheme for Core with Hexagonal Jacketed Fuel

Lou Lei, Wang Lianjie, Wei Yanqin, Huang Shien, Cai Yun, Chen Liang, Liu Xiaoli, Li Sinan, Tang Xiao, Zhang Ce

2021, 42(5): 256-260. doi: 10.13832/j.jnpe.2021.05.0256

Abstract(144) HTML (92) PDF(26)

Abstract:
This paper proposes 11 core critical physical test schemes and performs corresponding calculations, demonstration and analysis, based on the critical physical test contents of the core with hexagonal jacketed fuel, in order to verify the calculation accuracy and reliability of the nuclear design calculation codes CELL and CPLEV2 for such core. The critical mass measurement scheme considers a fine adjustment of the core layout in the presence of a deviation between the calculation and actual results to ensure that the deviation between the effective multiplication factor in the case of full rod withdrawal and critical state is acceptable. The demonstration shows that the core loading scheme proposed in this paper meets the reliability test requirements for the core design code and can be used as the critical physical test scheme for the core with hexagonal jacketed fuel.

Study on Optimization of Leaf Spring Characteristics Based on Particle Swarm Optimization Algorithm

He Daming, Li Yuanming, Pu Zengping, Wu Xingwen, Li Wei

2021, 42(5): 261-265. doi: 10.13832/j.jnpe.2021.05.0261

Abstract(255) HTML (127) PDF(32)

Abstract:
The leaf spring is an important part of a nuclear fuel assembly. Its performance is directly related to the service safety of fuel assemblies. Considering the complex non-linear factors of leaf spring materials, such as the elastic-plastic constitutive relation and large deformation, the authors, using the ANSYS code, performs the parametric modeling of leaf spring under the constraint of multi-scale coupling. The model geometry, hexahedral mesh and contact pairs of the leaf spring are established automatically. On this basis, the authors study and analyze the influence of key parameters on the characteristics of leaf springs. Depending on the MATLAB parallel computing library, the authors build a multi-parameter optimization platform for leaf spring characteristics based on intelligent particle swarm optimization (IPSO) algorithm. Then, the authors use this platform to optimize the leaf thickness, variable cross-section position and arc-shaped transition zone form over the design stiffness curve and minimum plastic deformation of the leaf spring. As demonstrated by the results, the intelligent multi-parameter optimization algorithm based on particle swarm can help improve the leaf spring design efficiency greatly; and within the given leaf spring design target curve and the leaf spring parameters, this algorithm can provide structural parameters satisfying the design target via a less times of iteration, which provide a good guidance on the nuclear reactor leaf spring engineering.

Study on the Concept of Organic-Cooled Microreactor

Li Qing, Xia Bangyang, Li Sinan, Lu Di

2021, 42(5): 266-270. doi: 10.13832/j.jnpe.2021.05.0266

Abstract(353) HTML (101) PDF(35)

Abstract:
In order to study the concept of organic-cooled reactor, this paper firstly analyzes the important characteristics and key technical problems of the organic fluid used as the reactor coolant and moderator, as well as the main technical schemes for organic-cooled reactors. On this basis, this paper studies the neutronic characteristics of the 5 MW microreactor core. As shown in the study, under the same core layout conditions, the absolute value of the moderator temperature coefficient of the organic-cooled reactor is below that of the pressurized water reactor (PWR), and the power distribution is more flat. The study results from this paper can provide an important reference for the determination of the technology roadmap for multi-purpose microreactor power supplies and heating systems, which are rapidly developed in China.

2021 Vol. 42, No. 5