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Research Progress of Metal-Based Shielding Materials for Neutron and Gamma Rays

Acta Metallurgica Sinica(English Letters) 2021年第12期34卷 1609-1617页

作者： Bin Gan Shichang Liu Zhen He Fucai Chen Haoxuan Niu Jingchang Cheng Bo Tan Bo Yu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu610213China State Key Laboratory of Light Alloy Casting Technology for High-End Equipment Shenyang Research Institute of Foundry Co.Ltd.Shenyang110022China

The radiation generated by nuclear reaction is harmful to human body and equipment,thus the radiation shielding materials that employ the shielding ability from neutron and gamma rays are the best candidates according to application situations and radiation *** this paper,the researches of metal-based neutron and gamma rays or multiple purpose shielding materials are systematically summarized,and the respective and principal problems of these materials with respect to shielding effectiveness and other performances,such as corrosion,mechanical properties,manufacture,etc.,are ***,the prospect of shielding materials is outlined,which suggests that the development of highly efficient and multiply functional radiation shielding materials with good environmental compatibility is one of the future development trends.

关键词： Radiation shielding materials Metal-based Boron-containing shielding materials High density alloys

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Sensitivity analysis for dynamical response of reactor coolant system based on optimus

Sensitivity analysis for dynamical response of reactor coola...

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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Yanli, Yuan Xianhui, Ye Lijuan, Li Feng, Yuan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791883785

The sensitivity analysis of the dynamical response of reactor coolant system to the input parameters is an important precondition for the design optimization. In this paper, the sensitivity of the dynamical loads at the nozzles of the equipment under seismic conditions is analyzed with an integrated platform called OPTIMUS, taking the stiffness of the dampers in the steam generator and the main pump as the input variables. The key parameters of the reactor system are usually different from the design value due to the calculation error, random and other uncontrollable errors in the manufacturing process and installation process. In a nuclear power project, the measured stiffness values of the dampers on the steam generator and the main pump in the manufacturer are deviated from the requirements in the equipment specification, and it is necessary to evaluate the influence of the deviation on the dynamical response analysis of the reactor system. According to the traditional method, it is necessary to establish the models of the reactor coolant system for nonlinear analysis according to the different stiffness of the dampers, and then the calculation results are compared by EXCEL. In this paper, the sensitivity analysis of output parameters which are the loads at the nozzles of the equipment to the input parameters which are the stiffness of the dampers on the steam generator and pump is realized by OPTIMUS, which is a kind of integration platform. Not only can ANSYS simulation calculations be carried out automatically on the OPTIMUS, but also the output data can be processed rapidly automatically, and the influence of manufacturing deviation of the stiffness of the dampers on the dynamical response of the reactor coolant system can be analyzed quantitatively in the above-mentioned problems, and the data support is provided for the determination of the design variables for subsequent optimization analysis. © 2020 American Society of Mechanical Engineers (ASME). All rig

关键词： Sensitivity analysis

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A NUMERICAL SIMULATION METHOD FOR THE VIBRATION REDUCTION OF MR ISOLATOR BASED ON IMPEDANCE TEST 30

A NUMERICAL SIMULATION METHOD FOR THE VIBRATION REDUCTION OF...

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30th International Congress on Sound and Vibration, ICSV 2024

作者： Wenzheng, Zhang Peng, Xiangfeng Zhihao, Yuan Guojiang, Liao Xiaozhou, Jiang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789090390581

It’s important to choose suitable method to simulate the vibration reduction of the isolators. For a long time, scholars and engineers have been working hard to reduce the error between numerical simulation and experimental results of the vibration level drop of the isolators. Due to the standing wave effect and the limitations of analytical models, it is difficult to simulate the vibration level drop of isolators well solely based on dynamic stiffness or damping. In this paper, an analysis model for the vibration reduction effect of metal-rubber (MR) vibration isolators based on impedance test data is established. The origin impedance and cross point impedance data is tested on the testing platform. The four-terminal parameter method is applied to calculate the stiffness and damping value of the MR vibration isolators. The calculation results show that compared with commonly used numerical simulation methods in the past, the model based on frequency dependent stiffness and damping can further reduce the error between the numerical simulation and the experimental results. © 2024 Proceedings of the International Congress on Sound and Vibration. All rights reserved.

关键词： Damping

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STUDY ON HYDROGEN RISK VENTING MITIGATION MEASURE OF SMALL STEEL CONTAINMENT DURING SEVERE ACCIDENT 29

STUDY ON HYDROGEN RISK VENTING MITIGATION MEASURE OF SMALL S...

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Zou, Zhiqiang Zhang, Hang Zhang, Ming Peng, Huanhuan Xu, Youyou Deng, Jian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886465

The hydrogen risk in small steel containment is one of the most important safety problems of small modular reactor (SMR), effective measures need to be taken to mitigate the risk of high concentration hydrogen in a small free volume during severe accidents. In the early stage of a severe accident caused by a loss of coolant accident, the pressure in the containment may rise rapidly, probably triggering the immediate venting of the containment, which can protect the integrity of the containment in the stage of pressure rise in the short term. In the later stage of a severe accident, the generation of hydrogen from reactor core may cause hydrogen combustion or even explosion, which may cause the possibility of overpressure in the containment. In this paper, Computational Fluid Dynamics (CFD) GASFLOW-MPI code is used to analyze the effect of containment venting measures on hydrogen risk in small steel containment of SMR during severe accident initiated by direct vessel injection line break loss-of-coolant. The analysis results show that the venting action can effectively depressurize the containment, the larger the area of the venting pipe is, the more obvious the depressurization effect is. However, hydrogen cannot be effectively discharged in the middle and later stage of the accident, so the risk of hydrogen cannot be completely eliminated, and the possibility of flame acceleration in containment cannot be ruled out for the reason that the negative pressure generated by water vapor condensation will suck in the air outside the containment and form a combustible gas cloud again. Copyright © 2022 by ASME.

关键词： Risk assessment

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SGTR ANALYSIS OFSMR ONCE-THROUGH STEAM GENERATOR 29

SGTR ANALYSIS OFSMR ONCE-THROUGH STEAM GENERATOR

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Zhu, Ye Xiang, Cheng Wei, Liu Zhiyun, Cai Yun, Ren Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886410

The steam generator tube rupture (SGTR) of a nuclear power plant is a transient that is subject to numerous studies. The casing once-through steam generator has great heat-transfer capacity due to its dual-layer heat transfer structure. Base on the structure and operational characteristic of casing once-through steam generator used in small module reactor (SMR), a graphical model is designed using process module in APROS. Operating parameters of the casing once-through steam generator is simulated under steady state of different power load and SGTR condition. According to the results, the enthalpy and the temperature of secondary side outlet first increases and then decreases with the increasing power load and reaches the greatest value when the power load is 60%. Further study the impact of the position and degree of the breakage on the operating characteristics of the casing once-through steam generator when the steam generator tube rupture. The results show that the influence on the operation characteristic of the casing once-through steam generator increase when the tube rupture takes place in the position which is near the primary side inlet. And when the size of tube rupture becomes larger, the leakage mass flow from primary side to secondary side increase, thus cause the enthalpy and the temperature of secondary outlet further decrease. The simulating results fit theoretical analysis, showing that the casing once-through steam generator can be simulated accurately in APROS. Reference and guidance can be provided for the design and optimization of casing once-through steam generator as well as the safety analysis. Copyright © 2022 by ASME.

关键词： Heat transfer

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Determination of Hardware Logic Drawing Simulation Modeling Criteria and Its Application in Nuclear Safety-Class DCS 6th

Determination of Hardware Logic Drawing Simulation Modeling ...

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6th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2021

作者： Zhang, Xu Chen, Shi-Yong Yao, Zhang Deng, Zhi-Guang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789811911804

In digital control system (DCS) design, independent hardware logic circuit design is usually used to deal with common causes failures of software. Due to the distinct difference between the design criteria of electrical design diagram and logic schematic diagram, the hardware logic verification should be carried out in the form of translation simulation. In this paper, the modeling criteria of DCS hardware logic simulation are summarized. The modeling criteria take into account both the standard engineering hardware logic drawing specifications and the characteristics of Simulink platform, and the DCS hardware logic electrical design information is translated to the Simulink platform for automatic redrawing according to specific criteria. A complete hardware logic simulation system was established based on the modeling criteria, and the application was verified in HPR 1000 Zhangzhou nuclear power plant project. © 2022, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Digital control systems

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Development and verification of SNTA code system for SCWR core steady state analysis 23

Development and verification of SNTA code system for SCWR co...

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23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015

作者： Wang, Lianjie Zhao, Wenbo Yang, Ping Ma, Yongqiang Lu, Di Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888982566

A coupled neutronics/thermal-hydraulics three dimensional code system SNTA is developed for SCWR core steady state analysis by modular coupling the improved neutronics nodal methodological code and SCWR thermal-hydraulic sub-channel code. The problem of CSR1000 core is studied to verify SNTA. The results calculated by SNTA are agreed well with those by CASIR and SRAC. SNTA is more efficient than CASIR and SRAC whose neutronics modules are based on the Finite Difference Method. The numeric results show that SNTA can be well applied to SCWR core steady state analysis and core concept design. Copyright © 2015 by JSME.

关键词： Finite difference method

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A revised method to simulate the dynamic characteristic of vibration isolator of main pump 27

A revised method to simulate the dynamic characteristic of v...

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27th International Congress on Sound and Vibration, ICSV 2021

作者： Zhang, Wenzheng Xiong, Furui Liu, Shuai Yuan, Zhihao Jiang, Xiaozhou Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9788378807995

It's important to choose suitable kinetic parameters for vibration isolators to simulate the dynamic characteristic. The kinetic parameters have significant influence on the vibration reduction of vibration isolators. In this paper, the kinetic parameters of two kinds of metal-rubber (MR) vibration isolators of main pump are measured to accurately acquire the dynamic characteristic. The origin impedance and cross point impedance data is tested on the testing platform. The four-terminal parameter method is applied to calculate the stiffness and damping value of the MR vibration isolators. Unfortunately, the method can't simulation the influence of the peak value behind the standing wave frequency, which is caused by the internal resonance of the vibration isolators. A revised rod vibration model is adopted to simulate the internal resonance, because the rod vibration model can express the high order peak. The numerical results are in good agreement with experimental results. The revised method can be used to simulate the dynamic characteristic of vibration isolators. © "Advances in Acoustics, Noise and Vibration - 2021" Proceedings of the 27th International Congress on Sound and Vibration, ICSV 2021. All rights reserved.

关键词： Kinetic parameters

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DATA DRIVEN METHODS FOR BREAK SIZE AND LOCATION ESTIMATION IN LOCA BASED ON DEEP LEARNING 29

DATA DRIVEN METHODS FOR BREAK SIZE AND LOCATION ESTIMATION I...

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Yuan, Peng Deng, Jian Qiu, Zhifang Zhu, Dahuan Huang, Tao Li, Zhongchun Du, Peng Xu, Qinglan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886427

Loss of coolant accident (LOCA) is one of the most important accidents in thermal hydraulic design of nuclear power plant. The traditional analysis method is difficult to realize the rapid prediction of the size and location of the break in the LOCA, while machine learning provides an idea for the rapid diagnosis of the initial cause of LOCA. In this paper HPR1000 which is independently designed by China was taken as the object, the process of LOCA under various break location and sizes is studied by using the advanced reactor system analysis code ARSAC (Advanced reactor system Analysis Code). The key thermal hydraulic parameters including temperature, pressure, flow rate and water level in the transient process of LOCA are analyzed, and a series of data sets are generated. Based on convolution neural network algorithm of deep learning, the transient thermal hydraulic parameters of HPR1000 under different working conditions are learned and the accident diagnosis model is obtained. By comparing with the test data, the trained model can quickly and accurately predict the size and location of LOCA, which can be subsequently extended to the diagnosis of other kinds of accidents in nuclear power plants and has certain application prospects. Copyright © 2022 by ASME.

关键词： Location

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COMPARISON OF NUMERICAL ANALYSIS METHODS OF COOLANT ENVIRONMENTAL FATIGUE BASED ON STRAIN RATE CONTROL 29

COMPARISON OF NUMERICAL ANALYSIS METHODS OF COOLANT ENVIRONM...

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Shao, Xuejiao Xie, Hai Zhang, Yixiong Xiong, Furui Wang, Xinjun Shi, Kaikai Yu, Mingda Huang, Xuan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886397

In this paper, the transformed strain rate of environmental fatigue correction factor (Fen), which is used to consider the influence of PWR primary coolant environment on material fatigue, is studied. EPRI's guideline and theAFCEN's probationary phase rule (RPP) No.3 of Volume VI in RCC-M 2017 edition provide two different calculation methods of transformed strain rate separately. NUREG CR-6909 gives two versions of prediction models of Fen for Low alloy steel, nickel base alloy and austenitic stainless steel which are the three main materials of the primary circuit system. In this paper, the model and transients of Sample 2 in EPRI's guideline are selected for calculation and comparison, so as to verify whether the analysis method is reasonable. Firstly, the fatigue analysis in air is carried out, and then based on the combination, the detailed calculation of strain rate is carried out to calculate environmental fatigue results. The analysis results are compared with the results of guideline, and the valuesare close. Then, the main pipeline safety injection nozzle components with small fatigue margin in PWR are selected to carry out environmental fatigue analysis. In the analysis, different methods of calculating modified strain rates are used, and different prediction models in CR-6909 are *** coefficient of Fenintegratedin RPP3 isconsidered, and the effects of various calculation methods on the results are compared. The results show that for austenitic stainless steel, the results of RCC-M RPP3 method are mostly less than those of EPRI guidelines. Compared with the prediction model of NUREG CR-6909 version 0, the prediction model version 1 greatly decreases the Fen results. Copyright © 2022 by ASME.

关键词： Strain rate

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