bluuebunny/crossref_metadata_2025 huggingface dataset:
Study on the Feasibility of Power Supply Migration in Production Preparation Area of the Nuclear Power Station
Authors: Yonggang Lin
Published: November 2022

Since 2016, Some nuclear power plants in the northern region has been facing severe situation due to seasonal heating; and because of seasonal hydropower generation, the generation situation of some nuclear power plants in the southern region is more serious. At present, the power supply in the production preparation area of most nuclear plants is mostly purchased electricity. There is a big difference between the price of electricity purchased from external power grid and the cost of power generation from nuclear plant. Therefore, if the production preparation area of the nuclear power plant power consumption is changed to power supply by itself (Auxiliary power), it will make a great contribution to energy saving and power saving, cost reduction and efficiency promotion. There is no study on the efficiency promotion of nuclear power generating units at home. This study mainly focuses on the project of power supply migration in production preparation area of nuclear power plant. This study is carried out from four aspects: transformer capacity, bus voltage drop of 6.6kV, capacitive current, and economy. The definition of production preparation area of nuclear power plant: the living area of the on-duty dormitory of the plant, administrative office area and other relevant areas.
Study on UC Phase Control of UC Ceramic Microspheres Prepared by Sol-Gel Method
Authors: Wen Feng , Guo Bolong , Li Mingyang , Wang Jiachen , Zhang Fan , Liu Wei , Ma Wenkai , Liu Zhihua
Published: November 2022

Since uranium carbide (UC) fuel has higher thermal conductivity and uranium density than UO2 fuel, good inherent safety and economy can be achieved by UC dispersion fuel. This paper focuses on the research of UC phase control during sol-gel preparation of UC ceramic microsphere. Using nano-carbon black powder as carbon source, UC ceramic microspheres were prepared by sol-gel process combined with carbothermal reaction process. Specific amount of carbon black was uniformly added into the glue solution, which was then dispersed to prepare carbon ADU gel spheres. With heat treatment of carbon ADU gel spheres subject to carbothermic reaction, ceramic microspheres containing UC phase were obtained. During the experiment, the effects of different carbon black addition ratio, thickener addition level and dispersion process on the chemical composition, sphericity and surface morphology on UC ceramic microspheres were investigated. The effects of heat treatment temperature, atmosphere and pressure on the preparation of UC microspheres were also studied during the carbothermic experiment. The process parameters for UC phase control were determined through the experiment and XRD, and the ceramic microsphere samples containing the UC phase were obtained as well. Analysis results show that crack-free UC microspheres with smooth surface can be obtained after drying when sol solution C/U is 2.5∼6.0 and thickener addition level is 1.2∼2.0 kg/kgU. By reaction treatment at high temperature, ceramic microspheres with good sphericity, compact structure and UC phase composition were obtained. A preparation method of UC ceramic microspheres was developed as a result.
Analysis of Single-Line Fast Starting Technology for Emergency Diesel Generators in Nuclear Power Plants
Authors: Li Hongzhou
Published: November 2022

This paper analyses the reasons for the failure of five consecutive single-line fast starting of emergency diesel generators in nuclear power plants. Based on the starting dynamics of diesel generators and governor control principles, and considering the starting characteristics of the respective units, the study identifies the root causes and key reasons affecting single-line fast starting among various contributing factors. The study extracted a set of methods for handing emergency diesel engine set faults. Employing a comparative analysis of test results and diagrams, the paper proposes solutions to ensuring the success of five consecutive single-line fast starting of emergency diesel generators for the first time in the world.
A Systematic Research on Prevention and Control Measures to Blockages of Cooling Water Intake System in Coastal Nuclear Power Plant
Authors: Cao Ranran , Yang Jia , Bai Wei
Published: November 2022

The cooling water intake system of coastal nuclear power plant involves multiple parts from the sea to the land. Moreover, the interception facilities of cooling water intake system are technically complex and face variable blocking risks. In order to improve the safety, stability and economy of nuclear power, for the first time, a systematic study on prevention and control measures to blockages of cooling water intake system in coastal nuclear power plant is carried out from the systematic perspective. The principle of active defense, depth setting, stepped filtration and permanent or temporary combination are adopted in this study. The research contents concern system design, equipment selection and technology layout, etc. Some key technologies, such as cold source blockages investigation, flow field active guidance and avoidance, cascade filtering system, pump station design, monitoring and early warning system, emergency response, etc., have been emphatically studied and optimized. As a result, a reasonable and systematic scheme of prevention and control measures to blockages of cooling water intake system is proposed. The application of this systematic scheme will enable nuclear power plant to form a strong-matrix prevention and control model containing all parts of the cooling water intake system. It can improve the active defense capability of cooling water intake system against blockages and reduce the occurrence of cooling water intake blockage incidents in nuclear power plant.
Thermal Experimental and Analysis of Vertical Spent Fuel Dry Storage Canister
Authors: Tang Qionghui , Li Mao-Yun , Cheng Cheng , Xie En-Fei
Published: November 2022

A full-scale prototype of vertical ventilated spent fuel dry storage canister is tested to evaluate its heat removal performance. Specifically, a temporary shelter is built to resemble the factory encompassing dry storage canisters. Electrical heating components with designed heat power are designed to represent the spent fuel assemblies inside the canister. From the test data its accordance with thermal safety related rules and guidance is proved. Furthermore, these data are compared with the simulated results, and the simulated results using CFD computer code of CFX. It is shown that the temperatures obtained from test agree well with the simulated results, and the simulated temperature of each component is no more than 10% higher than the test data, the velocity inside air-channel is in range of 1m/s, which is close to the environmental wind velocity of mid-summer of the testing site. It is also shown that conductive heat transfer vial air channel between the concrete over-pack and the steel canister covers more than 2/3 of the total heat load, indicating that conductive heat transfer plays a major role compared with heat conduction and radiation. From the full-scale prototype testing, it is shown that the vertical dry storage canister design has inherited conservatism from the aspect of thermal safety, And it is reasonable and acceptable to use CFD numerical calculation method for thermal analysis of this type of ventilated spent fuel dry storage system.
Reactor Pit Injection Isolation Design Modification and Experience Feedback of UK HPR1000 In-Vessel Retention
Authors: Zheng Hua , Wei Shuhong
Published: November 2022

During UK HPR1000 Generic Design Assessment (GDA), a shortfall about In-Vessel Retention (IVR) strategy is identified after a detailed timeline analysis according to UK nuclear safety requirements “Safety assessment principles for nuclear facilities” (SAP): the reactor pit injection manual action available time is less than the required time. Four possible modification schemes are proposed and analyzed. After ALARP (as low as reasonably practicable) analysis, the design of reactor pit injection isolation is modified to make the reactor pit injection manual action available time more than the required time : a remote manual permissive action in the main control room by a switch in the Emergency Control Panel replaces the local administrative isolation. Improvements for advanced nuclear power plant R&D, design and future operation are suggested, including enhanced requirement management and configuration management, especially for the design basis of functional requirements and sizing requirements; systematic identification of time-critical actions (TCA), performing analysis, preliminary validation and integrated validation of TCA during design and construction phase, implementing a TCA programme during operation phase; improving Severe Accident Management Guidelines (SAMG) implementation approval process by adding “the point of entry to the use of SAMGs is reached” to reduce the time of SAMG implementation approval process.
Experimental Investigation on Natural Circulation Characteristics of Core Make-Up Tank Under Rolling Conditions
Authors: Wang Shuguang , Zhang Jinlei , Shen Ya'ou , Liu Xiaobo , Ding Shuhua , Liu Yusheng , Li Dongyang , Tang Jilin
Published: November 2022

In the floating nuclear power plant, Core Make-up Tank (CMT) is one of the passive safety facilities, has been widely used in nuclear engineering. Under emergency conditions, natural circulation occurs and consequent thermal stratification appears in CMT system. However, rolling motion will directly affect the natural circulation and destroy the formation of thermal stratification, so there will be uncertain and unknown change of the safety injection capacity of CMT. Based on above analysis, this work carried out experiments under rolling conditions to reveal natural circulation and thermal stratification characteristics in CMT. A temperature detecting matrix formed by thermocouples is deployed in CMT. It is concluded that thermal stratification distribution is affected by the rolling conditions. Rolling conditions increase the thickness of thermal stratification and reduce the temperature gradient. The CMT natural circulation has two stages, the influence of rolling motion on the average flow in the first stage of natural circulation is very weak, but rolling motion will significantly affect the second stage of natural circulation and advance the decline time of direct vessel injection (DVI) flow rate. The transition time of natural circulation stage and the flow change in the second stage of natural circulation are related to the thermal stratification in CMT. The impact of rolling conditions should not be ignored when evaluating the performance of CMT.
Microstructure, Texture and Mechanical Properties in Different Deformation Regions of the Compressed Zr-4 Alloy With Lamellar Structure
Authors: Ren Jie , Li Geping , Yuan Fusen , Wang Ben , Han Fuzhou , Ali Muhammad , Zhang Yingdong , Guo Wenbin , Gao Bo
Published: November 2022

In the present work, the inhomogeneous plastic deformation of the compressed Zr-4 alloy with lamellar structure has been studied systematically. In the compression test of Zr-4 alloy with reduction of 70% at 650°C, the severe deformed area (SDA) and difficult deformed area (DDA) formed. The differences in DDA and SDA including microstructure, texture, and mechanical properties were investigated, using electron backscattering diffraction (EBSD) and Vickers hardness tester. The results showed that a large number of equiaxed grains appeared in SDA due to the occurrence of dynamic recrystallization (DRX), while coarse lamellar grains were still the main microstructure in DDA, showing that the DDA was the intermediate between the initial alloy and SDA. Furthermore, the discontinuous recrystallization mechanism was proposed. According to electron backscattered diffraction (EBSD) analysis, the predominant texture formed in SDA and DDA made normal direction of its {0001} planes of α phase parallel to the compression direction, and another stronger texture in SDA made the normal direction of part {0001} planes ∼15° away from the compression direction. The observed texture difference in SDA and DDA mainly resulted from the DRXed grains. To evaluate the influence of SDA on the mechanical properties of the Zr-4 alloy, microhardness tests were employed. The alloy in SDA had been proved to be strengthened compared to DDA, which was due to both the grain refinement and the texture diversity.
Optimizing Material/Component Fabrication Processing for Improving Service Performance in Nuclear Power Plant Coolants
Authors: Lu Zhanpeng
Published: November 2022

The service performances in terms of stress corrosion cracking (SCC) of pressurized water reactor structural components in coolants are critical for the economic and long-term safe operation. the materials, microstructual and mechanical properties resulting from material manufacturing and component fabrication are strongly affecting their performance behaviors such as environmentally assisted cracking. The integrated route for achieving high service performance materials and components is proposed. Some cases of reports on the plant experiences of using cracking-sensitive or cracking-resistant materials are used to illustrated the importance of the route. The route is practiced and demonstrated to be effective through our cooperative R&D works with designers, materials producers, component venders, operating service organizations as well as industrial research institutes.
A Two-Phase Flow Computation Code Based on Two-Fluid Seven-Equation Model
Authors: Zhang Hao , Yang Yanhua
Published: November 2022

The present work is developing a computation code based on two-fluid seven-equation model. This model is fully non-equilibrium. It not only assumes two phases are mechanical and thermal non-equilibrium, but also it assumes the pressure is non-equilibrium. The governing equations include seven partial differential equations, which are two continuity equations, two momentum equations, two energy equations, and one volume fraction evolution equation. The computation framework is based on split scheme, including Godunov split and Strang split, and the problem is split into two independent subproblems. One is a homogenous partial differential equation (PDE) problem, and the other one is an ordinary differential equation (ODE) problem. The homogenous PDE problem is reduced to linear equation by a Roe-type Riemann solver, and the linear equation is solved by upwind and MUSCL-Hancock (MHM) scheme. The ODE problem is solved by an implicit ODE solver, including Euler backward method, trapezium rule method and TR-BDF2. This code is verified by three types of problems. MHM converges faster than upwind for smooth solution problem, however MHM’s convergency rate is not always large than that of upwind for shock problem. The source term problem shows that the calculation result is sensitivity to the numerical schemes.
Study on Simulation Test of Hydrogen Deflagration Environment In Nuclear Power Plants
Authors: Zhan Huian , Jiang Shaokun , Tao Zhiyong , Zhao Ning , Yuan Shengnan
Published: November 2022

The safety control of hydrogen is a key safety issue that must be considered in the design of nuclear power plant. In order to better study the impact of hydrogen combustion and even deflagration on the nuclear power plant, the Handan Purification Equipment Research Institute has designed and built a hydrogen deflagration environment simulation test platform, which can simulate the deflagration process of mixed gas with high hydrogen concentration range. The test results show that there are some differences in the hydrogen deflagration process of test container with different volumes. The large volume test container is more in line with the change process of hydrogen deflagration environment in serious accidents of nuclear power plant. The large volume test container has problems such as large concentration gradient of hydrogen and temperature stratification. The closer to the top of the test container, the higher the temperature. It is possible to select a suitable location to install the test equipment according to the needs. With the same gas composition conditions, increasing the initial mixture temperature can effectively improve the half peak width of the hydrogen deflagration ambient temperature curve and slow down the temperature decline rate in the simulated environment. The above experimental research lays a foundation for the equipment availability verification test in the hydrogen deflagration environment of nuclear power plant.
Design Overview for Process System of Large-Scale Seawater Desalination and Demineralized Water Production in the Nuclear Heating Plant
Authors: Wang Rui , Xu Kechen , Zeng Kehao , Yan Hongmei , Bai Wei , Li Jing
Published: November 2022

A nuclear heating plant is a cogeneration nuclear heating plant, which provides industrial steam for the chemical industrial park to satisfy the steam demand of enterprises. Meanwhile, the remaining capacity can be used for power generation, so as to alleviate the local power shortage. As the condensed water of the externally supplied steam is not recovered, and the freshwater source is insufficient, it is necessary to build a seawater desalination plant and a demineralized water production plant to provide a large amount of demineralized water for the heating system. According to the raw seawater quality, requirements of produced water quality and technical & economic analysis, the main treatment process flow of seawater desalination system is: air flotation tank → horizontal mechanical filter → ultrafiltration membrane → seawater reverse osmosis membrane → freshwater reverse osmosis membrane. The main treatment process flow of demineralized water production system is: cation exchanger → anion exchanger → mixed ion exchanger. Combined with the engineering practice, this paper expounds the process system design of large-scale seawater desalination and demineralized water production from the aspects of seawater desalination process comparison and selection, reverse osmosis membrane seawater desalination process system design and process equipment configuration, etc. The system improves the guarantee rate of desalinated seawater and demineralized water supply, and realizes the emission reduction of industrial wastewater, which can provide reference for the process design of large-scale seawater desalination and demineralized water production in subsequent nuclear heating projects or nuclear power projects.
Research and Reflection on a Systematic and Integrated Application of Human Factors Engineering in Nuclear Power Plant
Authors: Xu Zhihui , Zhang Yuxin , Peng Huaqing , Yang Ming
Published: November 2022

Human factors engineering focuses on human-machine-environment coordination. The purpose is to prevent and reduce human errors in complex human-machine system, and improve the reliability of human-machine system. Based on the latest engineering practice of British generic design assessment for “HPR1000”, this paper introduces a systematic human factor analysis and argument technology by an integration of human factors engineering elements, as well as its engineering application. The engineering practice showed that this systematic method that integrates multiple elements of human factors engineering, can fully combine with the design characteristics of nuclear power plant, comprehensively analyze and consider the impact of human factors on nuclear power plant design, and can transmit and iterate risk information among various professional fields s. It is an effective way to realize the in-depth application of human factors engineering.
Experimental Study on the Feasibility of Dry Grinding of MOX Core Blocks
Authors: Mao Jianzhong , Liu Xin , Wang Ben , Zhang Pengfei , Gao Bo
Published: November 2022

The MOX core block is the main component of the MOX fuel assembly. In this paper, a resin-bonded diamond cup wheel was used for grinding experiments on three ceramic materials, ZrO2, CeO2+AL2O3 and CeO2+MgO, with similar properties to the MOX core block. A grinding temperature prediction model was established using the circular contact moving heat source model, and the grinding temperature of the workpiece surface was predicted and the control strategy was studied in conjunction with the grinding experimental results to evaluate the dry grinding adaptability of the MOX core block based on the workpiece surface grinding temperature. The results show that when the grinding wheel linear speed Vs = 60 m/s, the workpiece feed linear speed Vwt = 627mm/s, and the grinding depth is in the range of 0.01mm∼0.04mm single side grinding depth, the grinding temperature of the workpiece surface is below 200°C, and the oxidation temperature of the workpiece is not reached, therefore, the dry grinding method is considered suitable for the machining of MOX core blocks.
Preliminary Design on the Wide Dynamic and Fast Response Electronic System for the 4H-SiC PIN Fast Neutron Detector
Authors: Yong Ma , Shuhuan Liu , Shuangying Liu
Published: November 2022

Based on the measured dynamic range of fast neutron injection rate, 4H-SiC PIN detector fast neutron detection efficiency and detector output current signal distribution characteristics, this paper designs and develops a current sensitive front-end readout circuit based on a high bandwidth transimpedance amplifier (ADI LTC6268-10), analyzes the main factors affecting the front-end circuit characteristics, and explores the main methods to improve the linear range of the front-end circuit. The main methods to improve the linear range of the front-end circuit are explored. Second, according to the front-end circuit output signal characteristics, based on high-bandwidth ADC (AD9680) and Zynq UltraScale+ MPSoC XCZU15EG development board, a digital data acquisition system (DAQ) was designed and developed. The preliminary simulation and experimental test results show that the designed detector electronics circuit has a good linear response under the conditions of input current pulse frequency of 20MHz, peak value of 10μA∼100μA and pulse width of 5ns. The research results provide relevant technical support for the establishment of 4H-SiC detector electronics system.
Development and Verification of Heat Pipe and Fin Radiation Heat Dissipation Program Based on MOOSE Platform
Authors: Wang Sipeng , Chen Xiaoquan , Tian Rui
Published: November 2022

Nuclear energy is an integral part of the energy supply required for future space exploration as efficient energy. Although heat pipe reactor is the main development direction of space reactor. Radiation radiator quality accounts for a large proportion of heat pipe reactor quality. Therefore, it is essential to improve the heat dissipation performance of the heat pipe reactor, effectively reducing the quality of the radiator under the condition of ensuring that the radiation heat dissipation power of the radiator is not less than that of the same volume radiator, design the optimal heat pipe cooling scheme and promote the development of space heat pipe reactor. In this paper, based on the finite element multi-physical coupling MOOSE platform, a program is developed to simulate the physical field of the heat pipe reactor radiator. The simulation of the heat pipe, fin and other parts of the radiator end unit is realized, and the combination of heat pipe and fin calculation is completed. At the same time, the temperature field of the new heat pipe radiator unit is obtained. The simulation results for the heat pipe and fin module are verified by the temperature field calculation program and another multi-physical coupling software. Another multi-physical coupling software results are compared with MOOSE simulation results for the combination part. The results show that the MOOSE program can reasonably simulate the heat transfer state in the simulation of heat pipes, fins, and combinations. The temperature field agrees with the verification results.
Model Evaluation of Various Thermo-Physical Properties of Nanofluids and ANN Modelling for 10kWe Integrated Reactor
Authors: Zheng Ling-Yun , Zhang Zhi-Gang , Wang Xin-Wen
Published: November 2022

A 10kWe integrated reactor with Stirling generator is in design, to satisfy China’s power demand for both Earth orbit and deep space exploration in the next two decades. The integration of the core and the thermoelectric conversion system, reduces the number of tubes and pump structures, which leads to a higher energy conversion rate, fewer failure risks and coolant leaks. The waste heat of this reactor would be transferred through its heat pipes to its radiators, then to the space. Applying nanofluids would help reduce the heat pipe sizes, because nanofluids have great alterations in their thermo-physical properties with a small fraction of nanoparticles. Numerous models have been proposed to characterize the thermo-physical properties of nanofluids. However, it is found that researchers have different, sometimes even contradictory conclusions about some of the properties. At the same time, these properties could be affected by various aspects, the simple models are not sufficient for the reference. This work focuses on evaluating the models of density, specific heat capacity, thermal conductivity, viscosity, and Nusselt number of nanofluids with statistical methods, and provides reference thermo-physical properties for the design of the heat pipes in the space reactor. For this reason, a great amount of experimental data is collected. Profiting from the collected data, artificial neural network (ANN) models based on Pytorch are trained and compared with the other models.
Research on LOCA Dynamic Analysis Model of Reactor System
Authors: Liu Shuai
Published: November 2022

The reactor system has a large scale of components. And the system dynamic analysis model for loss of coolant accidents (LOCA) contains many non-linear factors. The transient calculation analysis takes a long time and the convergence is difficult to guarantee. In this paper, two different analysis models (complete three-loop model and single-loop model) of the reactor loop system are established for comparative analysis. The analysis shows that the calculation error of the main position load in the two different model is within 10%. The decoupling influence of the surge tube and the reactor inlet and outlet on the system dynamic characteristics is within an acceptable range. When the single-loop model is used, it is necessary to decouple the reactor pressure vessel and pressurizer. And the simplified boundary is adapted to simulate its effect on the whole loop. The single-loop dynamic analysis model based on the decoupling criterion can greatly reduce the calculation scale and meet the calculation accuracy, which can be used to calculate the LOCA dynamic characteristics of the reactor system quickly. The model can also be used for seismic analysis of reactor system. And using this model for reactor system dynamic analysis can greatly improve the calculation efficiency.
Research on the Practice Qualification System of Registered Nuclear Safety Engineer From the Perspective of Access Professional Qualification
Authors: Chen Fangqiang , Zhang Yang , Jia Wei , Li Zhaotong , Wang Guanyi
Published: November 2022

In order to improve the professional and technical level of China’s nuclear and radiation safety practitioners, enhance their safety literacy, China established Registered Nuclear Safety Engineer Qualification License System (hereinafter referred to as RNSES). Registered nuclear safety engineers not only need to master the knowledge of nuclear safety professional technology, but also need to master the knowledge of nuclear safety laws and regulations, department rules, nuclear safety guidelines, relevant national standards and industry standards, nuclear safety culture, etc. RNSES has good advantages and positive support for the implementation of nuclear safety regulations and national standards in the nuclear field. Nevertheless, the system has been established for 20 years and has accumulated some problems to be solved. Now, RNSES is planned to be adjusted. From the perspective of professional qualification of admittance, the paper makes a multi-dimensional statistical study on the construction of other professional qualification access systems, intuitively points out the realities and shortcomings of the registered nuclear safety engineer system, and then puts forward specific suggestions to strengthen the construction of the access system of certified nuclear safety engineers, so as to provide policy reference for the system reform.
Pulverization and Spheroidization of Natural Graphite Powder for The Matrix Graphite in High-Temperature Gas-Cooled Reactors
Authors: Cao Xinlei , Lu Zhenming , Feng Yuxiao , Bai Juan , Liu Bing , Shen Ke
Published: November 2022

Matrix graphite (MG) constitutes a major component of the fuel element in high-temperature gas-cooled reactors (HTRs), and its performance directly affects the economy and safety of HTRs. A3-3 MG, developed in Germany in the last century, is still used today without significant changes. However, with the successful commercialization of HTR, there is an urgent need to develop MG that is suitable for the characteristics of each country. In this paper, comparative studies on the powder processing of natural flake graphite were carried out based on the raw materials from Inner Mongolia. First, the influence of the sequence of purification and pulverization on the basic properties of graphite powder was studied. Second, the influence of the degree of spheroidization of the natural flake graphite powder on the properties of the MG was further investigated, and the controlling mechanism of the spheroidization process on the microstructure and physical properties of the MG was revealed. The results of this study provide important insights for optimizing the preparation process of HTR fuel element MG.
Research on Laser Welding Process and Deformation Control of Irradiated Sample Holder
Authors: Luo Shouyi , Guo Baochao , Mi Dawei
Published: November 2022

In order to study the influence of laser welding parameters on the weld formation and the deformation of the irradiated sample holder, laser lap welding tests were carried out on 304 stainless steel sheets with the thicknesses of 8.97mm and 20mm. Firstly, orthogonal tests were carried out to study the effect of welding speed and defocus on weld morphologies. The results showed that when the laser power was 10kW, the welding speed was 15mm/min, and the defocus amount was −1mm, the optimized weld with the aspect ratio of 2.13 can be obtained. Next, a novel fixture designed for the welding of the irradiated sample holder was introduced, which effectively reduced the deformation of workpieces during the welding process. Additionally, in order to compensate the deformation induced by the weld shrinkage and residual stress, a filler gauge of 0.75mm height was used to create an inverse deformation. it was proved that the welding deformation could be effectively controlled after the preforming, and the deformation could be controlled below 0.7mm. Finally, the influence of welding sequence on the deformation of the irradiated sample holder under the optimized welding parameters was investigated. It was shown that the optimal welding sequence of the irradiated sample holder was: firstly the arc positioning welding, then the laser positioning welding, and finally the laser deep penetration welding. The welding deformation can be effectively controlled by adding the fixture and adjusting the welding sequence, the change of flatness before and after the laser deep penetration welding could reach 1.09mm.
Simulation Research on Safe Load-Reducing Operation Margin of Nuclear Power Unit With Clogging of the CFI Drum Net
Authors: Liu Jinchao , Lin Yaozu , Chen Bijun , Lyu Xue , Li Jiayi , Zhang Heng
Published: November 2022

As the final heat sink of the nuclear power plant, the cool source is responsible for the safety function of the waste heat export of the reactor. The explosive invasion of marine organisms and other wastes have a great impact on the cold source system of nuclear power plants, which will lead to transient events, load reduction and shutdown of nuclear power unit, and even affect the availability of the cold source system in severe cases. which are huge threats and challenges for the safe and stable operation of unit. thereby reducing the power generation efficiency, safety, economy, and sustainable development of nuclear power plants. Clogging of the CFI drum net is one of the important factors causing cold source system accidents. this paper puts forward the solution idea of maintaining the water level of the forebay through the variable frequency pump, so as to ensure that the CRF pump keeps running state under the condition of cold source blockage, so as to maintain the function of the cold source system. and based on the full range simulator of nuclear power units, consider CFI drum network differential pressure high alarm failure combined with a monitoring device for the blockage degree of the drum net, the scheme is simulated and verified, and the load that the nuclear power unit can operate safely under different blocking degrees of drum network is given. the research results of this paper provide a solution for alleviating the cold source problem, and the solution can also provide a reference for the safety improvement of the cold source of actual unit and the design of new reactor.
3-D Radiation Field Reconstruction for a Facility With Multiple Radioactive Sources
Authors: Zhu Shangzhen , Dong Xinwen , Zhuang Shuhan , Fang Sheng , Cao Jianzhu , Li Wenqian
Published: November 2022

The radiation field measurement and surveying play important roles in optimizing and planning operation work in the radioactive area. In this work, a 3-D radiation field reconstruction method is applied to reconstruct a 3-D radiation field of a facility with two radioactive drums and a shielding, not limited to the existing methods under two-dimensional grids. A numerical scheme of the method is introduced. Monte Carlo simulation of the gamma radiation field was performed to be an original field so that the reconstruction results can be verified. Random and regular sampling ways are taken into consideration while the sampling rates are kept at 4.12%. Quantitative evaluation of the results evaluates MSE (mean relative error) and MRE (mean squared error), MRE is kept at less than 4%, which shows a good reconstruction accuracy of the method in almost all 3D space. The work in this paper has good reference value for applications such as radiation field detection, inversion, and reconstruction, operations under radioactive environments like nuclear waste decommissioning, and nuclear power radiation modeling.
Simulation of Neutron and Proton Displacement Damage in Ultra-Wide Bandgap Semiconductor Ga2O3
Authors: Shao Zhuang , Cai Ziqi , Zhang Qingmin
Published: November 2022

Recently, gallium oxide (Ga2O3) as an ultra-wide bandgap oxide semiconductor has aroused enormous attention in research and development due to its prospects for future power electronic, optoelectronic, and radiation detection applications. Ga2O3-based devices could be subject to fluxes of protons or neutrons if used in aerospace or radiation-hard nuclear systems, which leads to internal defects in Ga2O3 crystals and the degradation of device performance. Therefore, the radiation effects of Ga2O3 under irradiation have become a research focus, and it is of great significance to study the defect behavior and performance influence of Ga2O3 after irradiation. In this paper, the number of displacement defects produced by 1∼20 MeV neutrons in Ga2O3 were calculated using Geant4 simulations, and the factors that impact the displacement damage were studied. The results show that the displacement defects generated by neutrons do not increase monotonously with neutron energy but are closely related to the reaction cross-section and the generation of (PKA). We simulated and calculated the radiation damage by 10∼100keV protons in Ga2O3 using SRIM. It is found that ionization damage is much greater than displacement damage; the number of vacancies generated by proton radiation in Ga2O3 increases with the energy and incident angle of the incident proton. The irradiation resistance of Ga2O3 is between silicon and diamond semiconductor materials.
Neutronics Analysis of Fusion Blanket Based on the Spherical Harmonic Function and Finite Element Method
Authors: Li Kang , Cao Liangzhi , Miao Jianxin , Zhang Haoyu , Dai Tao
Published: November 2022

The fusion neutronics simulation has been a great challenge to the numerical calculation of the neutron-transport equation. The key is how to deal with the features of fusion devices, such as large-scale, complex geometric models, large vacuum regions, etc. NECP-FISH, a code developed by Nuclear Engineering Computational Physics (NECP) laboratory of Xi'an Jiaotong University, is used to address this challenge. NECP-FISH adopts a deterministic numerical method instead of the Monte Carlo method because the deterministic numerical method is of higher computational efficiency and costs less computational time. To deal with large vacuum region, large-scale and complex geometric model, the first order neutron-transport equation is solved, the spherical harmonics function and the finite element method are applied to the expansion of angle and space. NECP-FISH has been validated by benchmark problems such as strong absorption problem, internal void problem, and Kobayashi series of problems. What’s more, NECP-FISH builds the user interface based on the platform SALOME so that users can visually build the necessary models for problems. NECP-FISH has been applied to the neutronics calculation of the breeder unit of Helium Cooling Ceramic Breeder (HCCB) and the blanket of CFETR. The numerical results demonstrate that the NECP-FISH code can efficiently solve the neutron transport problem of the fusion reactor.
Research on Technical Scheme of Rapid Power Reduction System Based on Gray Control Rods
Authors: Li Wei , Shi Jianfeng , Qin Yulong , Wang Lihua
Published: November 2022

The paper designed a small rapid power reduction system (SRPRS) based on gray control rods for pressurized nuclear power plants, and technical scheme of SRPRS was proposed in the paper. Different to current RPRS for turbine trip, SRPRS was designed for feed water pump trip. When one FWP happened, partial of gray banks dropped into the reactor core to reduce the core power, and reactor trip due to power mismatch between primary and secondary sides would be avoided. The paper researched the basic principles of SRPRS, and demonstrated the feasibility of using SRPRS in pressurized nuclear power plants with gray control rods. The small rapid power reduction system designed in the paper includes three stages: activation, recovery and normal operation. Each stage can automatically complete control actions with a few human intervention. The small rapid power reduction system makes full use of the gray control rod characteristics: low reactivity worth and small disturbance to core power distribution, so it can achieve rapid reduction of core power while maintaining a flat power distribution. Meanwhile, since the control rod insertion limit set in the technical specification does not limit the insertion of gray control rods, the operation of small rapid power reduction system will not exceed the requirements of the technical specification for normal operation. The paper designed a “slope factor” in small rapid power reduction system. By calculating the slope factor in advance and judging the rods to be dropped in real time, the small rapid power reduction system can perform its function at any initial gray control rod position.
Determination of Control Rod Position by Global Optimization Methods
Authors: Qian Zhang , Liu Zhihong , Zhao Jing
Published: November 2022

The method is developed to find the implicit position of the control rod using the flux value shown by the neutron detectors. When using the 3D core physics analysis code CPACT, the detector response of a set of predicted rod positions can be calculated. On a double duel problem, functions can be built to describe the difference between the prediction and the target. Under the frame of the inverse problem, combinatorial optimization methods are operated to renew the prediction of rod positions by iteration to a result with convergence. By such a method, the results are guaranteed to achieve the self-consistent with the physical and thermal field while reaching stable convergence when the forward calculation software has been fully verified to simulate the real case. In the meanwhile, the appropriate computational cost is needed and can be adjusted with time tolerance. So, we have compared several global search algorithms (Genetic Algorithm, Particle Swarm Optimization, and Characteristic Statistic Algorithm) with their performance. We can also get the conclusion that hidden features and intermediate calculation results are important to improve the search efficiency. However, the responses of the detectors to some remote rods are not obvious, which are not easy to determine. All the methods were tested on the numerical calculation examples of a pressurized water reactor case.
Study on Ultrasonic Inspection Technology of Fastening Bolts of Reactor Internals
Authors: Tang Jianbang , Wu Yukun , Wang Weiqiang , Ma Guanbing , Zeng Chenming , Xu Junlong
Published: November 2022

Reactor internals are the core components of the reactor pressure vessels in Nuclear Power Plant (NPP), which directly compress, support and position the nuclear fuel assemblies. They are exposed to high temperature, high pressure and high radiation during a long-term service. The fastening bolt may break and fall off due to irradiation stress corrosion crack (IASCC), which directly affects the safe operation of the reactor. Along with the in service time increase, it has gradually become the focus of the in-service supervision of NPP, and it is urgent to perform Nondestructive Examination (NDE) for its structural integrity. In this study, according to the structural characteristics and inspection difficulties of typical bolts, a special ultrasonic combined probe and ultrasonic process are designed, by using simulation technology. The design principles are summarized. In addition, the automatic inspection system of underwater submarine is introduced. The results of the process verification show that currently technology can effectively detect with a minimum flaw sizes of 17.5% cross-sectional area (CSA), which meets the in-service inspection (ISI) requirements of NPP.
Regulation without Transparency Impedes Compliance
Authors: Miranda Samuel
Published: November 2022

In November 2020, the Office of the Inspector General (OIG) of the Nuclear Regulatory Commission (NRC) issued its annual audit plan, which includes an audit concerning transparency (i.e., the public trust and confidence) in the NRC’s regulatory activities. [3] The audit will examine the NRC’s practice of allowing “drop-in” visits, by senior executives of licensees, especially during times when the NRC staff could be evaluating and deciding on regulatory matters that directly affect the interests of those licensees.

In 2017, the OIG conducted a related audit that also focused upon the public’s trust and confidence in the NRC. [4] That audit examined the NRC’s procedure that governed its evaluation of 10 CFR §2.206 enforcement petitions. The OIG found that the NRC staff had not issued a single enforcement order, as the result of 38 enforcement petitions that it had received in the prior three fiscal years, ending in 2016. The OIG concluded that, the lack of such actions could adversely affect the public’s perspective on the effectiveness of the agency’s 10 CFR 2.206 petition process. Both audits are discussed herein, via examples that illustrate the NRC’s implementation of its policy of transparency, in theory and practice.
Numerical Study of the Effect of δ-Ferrite on Crack-Tip Mechanical Field
Authors: Wu Yule , Zhang Lefu
Published: November 2022

A dual-phase microstructure with 5–12% δ-ferrite is needed to prevent hot cracking of stainless steel weld metal. However, the δ-ferrite also makes it susceptible to long-term thermal aging embrittlement. Recent studies revealed that the stress corrosion cracking (SCC) susceptibility might either be reduced or increased by the δ-ferrite, depending on the embrittlement degree of the ferrite during the operation. One possible explanation was that the harder δ-ferrite inhibited the crack growth by changing the cracking direction and creating a highly branched crack path. To further reveal the mechanism of δ-ferrite effect on SCC behavior of those materials in high temperature water, a finite element investigation for the distribution of crack-tip stress and plastic strain field with and without δ-ferrite under a constant load was conducted. The result shows that the crack tip stress and plastic strain decrease when the crack encounters the δ-ferrite, indicating a lower cracking susceptibility, and the enhancement of δ-ferrite hardness can further reduce the crack-tip plastic strain. the severe stress and strain states occur along the ferrite edge, this indicates that the cracks are more tend to initiate and develop along the interface of δ-ferrite and austenite matrix.
Discussion of Nuclear and Radiation Safety Regulator Training in China
Authors: Li Zhaotong , Xing Dan , Chen Fangqiang , Jia Wei , Zhang Yang , Chen Shuai
Published: November 2022

China’s National Nuclear Safety Administration (NNSA) has always attached great importance to the training of regulators. In 2020, NNSA issued the training program for nuclear and radiation safety regulators. In this study, research was conducted on the implementation of the training program. Analyze the current status of our training systems were drawn from the aspect of implementation and management. Main problems of regulators’ systems were summarized from this study: (1) the training course needs to be further improved according to the training program. (2) The construction of curriculum materials needs to be continuously promoted. (3) personnel training experience needs to be systematically managed. (4) training knowledge management and application need to explore effective ways. Solutions to each problem are proposed to help improve the training efficiency of regulators and explore the use of training knowledge.
Research on Fatigue Analysis Method of Nuclear Equipment Under Ocean Conditions
Authors: Liu Guangdong , Zhu He , Zhang Zhui , Zhang Guoxun , Cao Guitao , Tan Lin , Qi Jing
Published: November 2022

Under the conditions of the marine environment (wind, wave, current), the floating platform on the ocean undergoes heeling, trimming, rolling, pitching, swaying, surging, heaving movements. The floating reactor on the floating platform within the design life, in addition to withstand the low-frequency fatigue caused by primary loop temperature and pressure, it also has to withstand the inertial force generated by the sway of the wave for the service period. The inertial force generated by the wave is a constantly changing dynamic load. Alternating stresses inside of the structure may cause fatigue damage to the structure. In order to estimate the fatigue life of floating reactor equipment under various ocean conditions, a finite element analysis model of equipment was established to carry out research for fatigue analysis methods under sway loads, meanwhile the operating temperature and pressure loads were considered. The probability of sway at each angle obeys the Rayleigh distribution, through the analysis of the status for the platform, it can estimate the number of swings at each angle. Based on Miner linear cumulative damage theory, the cumulative fatigue usage factors meet the requirements of the ASME BPVC III NB-3220 code. This analysis method provides a reference for the fatigue analysis of the floating reactor nuclear equipment.