bluuebunny/crossref_metadata_2025 huggingface dataset:
Study on the Influencing Factors of Oxidation Degradation for High Concentration Polyvinyl Alcohol Wastewater in Heterogeneous Fenton System
Authors: He Ye , Zhou Dongsheng , Liu Feng , Li Xuefei , Su Xingdong , Li Jiawen , Zhan Jie
Published: November 2022

Polyvinyl alcohol (PVA), a high molecular polymer, has good barrier property, acid or alkali resistance and degradability, which could be used as personal radiation protective equipment and pollution control articles in the nuclear industry, resulting in degradable radioactive technical wastes after use. The degradation treatment process of PVA degradable technology waste can realize validity separation of radionuclides and reduce the output of radioactive solid waste. In this study, MnxOy/γ-Al2O3, CuxOy/γ-Al2O3, FexOy/γ-Al2O3 and FexCuyMnzOw/γ-Al2O3 nanoparticles were prepared by the impregnation-calcination method, which were used as catalysts for the oxidative degradation of ultra-high concentration PVA organic waste solution, and the effects of initial pH of polyvinyl alcohol (PVA) solution, hydrogen peroxide (H2O2) dosing, catalyst dosing and initial concentration of PVA solution on the degradation of PVA were analyzed. The results showed that the best overall degradability was achieved by using FexCuyMnzOw/γ-Al2O3 nanoparticles as catalysts with the initial reaction concentration of PVA at 1%, the initial pH of the solution at 3, the reaction temperature at 60 °C, the catalyst dosage at 0.5 g/L and the H2O2 dosage at 60 mL/L. At this time, the concentration decrease rate of PVA could reach more than 95 %, the viscosity average molecular weight of degradation products reached 1395, the molecular weight decrease rate reached 99 %, and the degradation products mainly included butanone, benzaldehyde, hexanal, acetic acid and other unsaturated small molecules. In addition, the mechanism of hydroxyl radical generation by nanoparticles catalyzing hydrogen peroxide was discussed, and the reaction pathway of oxidative degradation of PVA macromolecules by hydroxyl radicals was proposed, which provided a reference for volume reduction treatment of PVA degradable waste in nuclear industry.
Analysis of Radiation-Induced Demagnetization Influence on the Performance of Permanent Magnet Damper
Authors: Wu Hongyu , Yan He , Diao Xingzhong
Published: November 2022

The magnetic vortex damper, a key unit of the control rod drive mechanism of high-temperature gas-cooled reactor (HTR), is a permanent magnet damper that produces eddy current resistance by rotating the conductor disc in magnetic field, and acts as a speed limit during the rod dropping process. To analyze the influence of neutron irradiation leaking from the core on the damping performance of the permanent magnet damper, the mechanism and the influencing factors of radiation-induced demagnetization of the Nd-Fe-B magnet were summarized through literature investigation and a magnetic vortex damper simulation model was established based on ANSYS Maxwell software and verified by experimental data. Current research shows that the magnet with higher intrinsic coercivity and length diameter ratio results in less demagnetization. The magnet permeance coefficient was simulated by the static magnetic field simulation to calculate its equivalent length diameter ratio. According to the literature experimental results, the equivalent length diameter ratio of the magnet, and its intrinsic coercivity, it was conservatively estimated that the remanence attenuation amplitude of the permanent magnet in this study should not exceed 1%. Based on this simulation model, the damping torque before and after the certain amplitude of magnet remanence attenuation was simulated and the corresponding maximum rod dropping speed was calculated. The simulation results show that the damping torque decrease is within 5%, which can meet the service requirements of high-temperature gas-cooled reactor over its service life.
Establishment of Engineering Economics Standard System for Nuclear Power Plants in China
Authors: Yang Wen , Zhang Shaolin , Huang Jun , Wu Feifei
Published: November 2022

Standard system has been developed for Nuclear Power Plants in China during the last ten years, which covers the entire lifespan of pressurized water reactors. It is mainly composed of the following parts, including general & foundation, preliminary work, designing, equipment, construction, commissioning, operation and decommissioning. Engineering economics is an important part of general & foundation section. Standards related to cost and cost management constitute engineering economics standard system, which is one of the most important subsystems in the standard system for Chinese nuclear power plants. In general, engineering economics standard system is composed of 39 nuclear industry standards, with 4 management standards, 6 specifications of bill of quantities and 29 cost quota standards.

Management standard specifies the engineering management and engineering cost evaluation for the construction of nuclear power plants. For the technical part of engineering economics standard system, two kinds of pricing methods for construction projects in china should be considered: fixed quota pricing and bill of quantities, and those standards should cover the overall construction content and construction phase of nuclear power plants.

This paper presents the establishment of engineering economics standard system framework and major elements in this system. The study in this paper shows a good example of the establishment of the standard system in engineering economics field for nuclear power plants in China.
Effect of a Cr2O3/A-Al2O3 Coating on the Corrosion Behavior of Incoloy 800H Alloy in High Temperature Impure Helium
Authors: Li Haoxiang , Zheng Wei , Du Bin , Yin Huaqiang , Hong Zhihao , Wang Long , He Xuedong , Ma Tao , Yang Xingtuan
Published: November 2022

In high temperature gas cooled reactor (HTGR), there is a small amount of helium in the primary coolant (ppm level) impure impurities, which will corrode the alloy materials of high temperature reactor and affect the performance of equipment in high temperature environment. The coating preparation technology on the alloy surface can effectively protect the alloy matrix, greatly reduce the corrosion rate of alloy, and improve the reliability and safety of equipment. In this paper, a Cr2O3/α-Al2O3 multilayer barrier coating prepared by RF magnetron sputtering was used to coat Incoloy 800H alloy, and the corrosion behavior of Incoloy 800H alloy with and without coating under selected high-temperature impure helium was investigated. After the experiment, the corrosion results were analyzed by weighing, scanning electron microscope (SEM) and X-ray energy dispersive spectroscopy (EDS). The results showed that Cr2O3/ α-Al2O3 barrier coating partially fell off after high temperature corrosion; The internal oxidation phenomenon of the alloy was not obvious, and the coating can inhibit the oxidation and carburization of the alloy; The mass of the alloy increased after corrosion, and the mass gain of the coated alloy was lower than that of the uncoated alloy.
Analysis on the Effects of Thermal Migration Phenomena to the Fuel Performance in the Fast Reactor
Authors: Yang Guangliang , Wang Weixiang , Ding Tao , Chen Hongli
Published: November 2022

Thermal migration phenomena in the fast reactor can affect the temperature distribution and in-pile behaviors heavily through changing the local properties of nuclear fuels. In this research, the classical thermal diffusion equations for the oxygen diffusion, plutonium migration, Joint Oxide Gain (JOG) formation, and porosity migration are solved and integrated into the fuel rod performance analysis code KMC-Fueltra as thermal migration analysis module. Validations of these models have been performed by comparing with the experimental data or simulation results. Comparisons have proved the correctness of the thermal migration module in KMC-Fueltra. Effects of these phenomena to the fuel rod performance are analyzed and discussed under the steady-state operating conditions in the 1000MWth Medium-size Modular Lead-cooled Fast Reactor (M2LFR-1000) using multi-physics simulation of coupled thermal analysis, thermal migration analysis, fission gas release analysis and mechanical analysis. Simulation results show that the influence of thermal migration mainly reflects in the fuel pellet other than the cladding. It can result in large changes in some parameters during the evaluation of the fuel rod performance, compared with simulations that do not account for thermal migration phenomena.
The Drafting and Discussion of Terminology of Aging and Life Management for Nuclear Power Plant in China
Authors: Zhong Zhimin
Published: November 2022

From 2016, we started to draft the industry standard of “Terminology of aging and life management for nuclear power plant “(NB/T 200555-2019)”. After three years, it was issued in China and came into recommend implementation at July, 2020. As ageing management and life management activities are important part of periodic safety review or operating license renewal for nuclear power plants, the standardization of ageing management and life management terms and terminology will benefit the related practice, technical communication and international cooperation. This paper briefly introduced the drafting background, aims, principals, methodology, and process of this industry standard in China. The standard’s scope, structure and terms category were discussed and analyzed briefly also. Then the difference between it and EPRI “Nuclear Power Plant Common Aging Terminology” (TR-100844) was discussed. After that, polishing the definitions of some essential terms was introduced in detail. Some of drafting experience was discussed also. In final, the issued standard includes 124 terms, which be classified into five categories as basic concept, aging mechanism, aging effects, aging management and life management. We believed that it will contribute the standardization and application of relevant terminology and benefit international communication in nuclear power plant aging and life management area.
Experimentally Investigating Effects of Water Quality on the Critical Heat Flux for the Downward Facing Surface
Authors: Zhang Guanhong , Liang Junming , Yun Xiangyu , Zhang Huiyong
Published: November 2022

In Vessel Retention (IVR) is an important severe accident management strategy and the critical heat flux (CHF) is an important factor for the reliability of the IVR. CHF refers to the heat flux that occurs before a boiling crisis with a sudden jump in surface temperature or a sharp deterioration in heat transfer rate.

In this paper, a small scale pool boiling facility is established to investigate the heat transfer coefficients of the downward facing surface under different water quality (including deionized water, tap water and seawater). Steady-state experiments were performed to research the effects of cooling water quality on the heat transfer coefficient.

The pool boiling experiments are finished under different power input with three kinds of fluid. The results show that: 1) There was an obvious boiling conversion process between deionized water and tap water, while there was no obvious boiling conversion process in seawater and the heat flux was significantly higher than the others. 2) After the experiment in seawater, the surface of the experimental section changed obviously due to the scale formation of seawater.
Software Design of High Reliable Decentralized Control System
Authors: Zhou Shuqiao , Li Duo , Guo Chao , Chen Fan , Huang Xiaojin
Published: November 2022

Digital instrumentation and control (I&C) systems are widely used in many industrial areas. In the recent years, the digitalization process for nuclear power plants has also been moving on rapidly. Full digitalized I&C systems are now adopted in almost all new constructed nuclear power plants. Moreover, for the advanced small modular reactors, both the high reliability and extendibility of I&C systems are especially required. The original architecture, which usually consists of centralized control cabinets and severs, of I&C systems cannot adequately provide the required extendibility. A novel hardware architecture of distributed control system aiming to enhance the reliability and extensibility was proposed and presented in the 26th International Conference on Nuclear Engineering (ICONE26). In this paper, we further focus on the considerations about the software design of this novel decentralized architecture. The hardware entities and their characters related to the software design are first introduced. Then the key considerations, including the design objective and software framework are described. Moreover, the related algorithms regarding the coordinate node selection and task assignment are proposed and demonstrated. At last the analysis about the software for the decentralized architecture are provided. The previously proposed decentralized architecture becomes feasible based on the software design introduced in this paper.
Analysis on the Fastening System of Radioactive Material Packages In Railway Transportation
Authors: Sun Qian , Wang Zhipeng , Wang Changwu , Zhang Yuhang , Chen Lei , Wang Xuexin , Li Guoqiang
Published: November 2022

For radioactive material packages railway transportation, a reliable fastening system of the packages in the wagon is the key to realize safe and efficient transportation. The following types of fastening systems are mainly used in the railway transportation for small radioactive material packages, such as stopper reinforcement, rope reinforcement, bolt reinforcement, freight container loading and reinforcement, etc. The fastening system should be determined according to the size, shape, weight and other parameters of different packages. In that case, the requirements of convenient operation could be considered while ensuring the reinforcement strength. Combined with the examples of different types of radioactive material packages, some examples of reasonable fastening systems for railway transportation were introduced. In order to ensure the safety, the strength of reinforcement materials should be calculated and verified, and the transportation stability of wagon loaded with packages should be analyzed.
Extraction of Nd(III), Eu(III), Am(III) and Cm(III) With 6-Carboxylic Di(2-Ethylhexyl) Amide Pyridine
Authors: Xu Chao , Du Yu , Liu Tingting , Yang Suliang
Published: November 2022

Solvent extraction has been widely used in spent fuel reprocessing because of its advantages such as high mass transfer rate, short production cycle, easy operation and large extraction capacity. The ligands containing soft S and N atoms usually have a good effect on the separation of trivalent lanthanides actinides. Herein, a novel extractant, 6-carboxylic di(2-ethylhexyl) amide pyridine (DEHAPA, HA), containing carboxyl and amide pyridine, was designed. The extraction of Nd(III), Eu(III), Am(III) and Cm(III), representing trivalent lanthanides and actinides, from nitric solution has been carried out by DEHAPA diluted in toluene as the organic phase. According to the slope analysis, the results show that the extraction of Ln(III) and An(III) with DEHAPA was governed by ion-exchange mechanism and the extraction equilibrium constants of Nd(III), Eu(III), Am(III) and Cm(III) have been calculated. The effect of concentration indicated that the structure of extraction complexes are 1:3/LnA3 and 1:3/AnA3. The temperature has a slight influence to distribution ratio of extraction Nd(III) and Eu(III). The infrared spectrum of DEHAPA and extracted complex analysis showed that -N-C = O and -O-C = O group coordinated with Nd(III). According to 1:3/LnA3 extracted complex structure, the Nd(III) ion in complex was coordinated with three -N-C = O, -O-C = O and pyridine group from three tridentate A− ligands by three tridentate A− ligand in organic solvent. This work reveals the unique extraction and coordination structure and provides a value reference to design more effective extraction ligands for Ln(III)/An(III) separation.
Prediction of the Operating Control Rod Position of the HTTR With Supervised Machine Learning
Authors: Ho Hai Quan , Nagasumi Satoru , Shimazaki Yosuke , Ishii Toshiaki , Iigaki Kazuhiko , Goto Minoru , Simanullang Irwan Liapto , Fujimoto Nozomu , Ishitsuka Etsuo
Published: November 2022

During operation of the HTTR, hundreds of technical signals and operating conditions must be observed and evaluated to ensure safe operation of the reactor. The accumulated experiment data of the HTTR is not only important for the HTTR operation, but also for the basic development of the HTGR in general. Artificial intelligence (AI) and particularly machine learning (ML) could give the ability to make predictions as well as allow the extraction of key information about physical process from large datasets. Hence, there is a lot of potential to apply AI and ML to predict the operating and safety parameters of the HTTR.

In this study, the control rod position of the HTTR is predicted based on ML without using the conventional neutronic codes. The ML with a linear regression algorithm finds a functional relationship between the input dataset and a reference dataset, constructing a function that predicts control rod position from the other operation conditions. As result, the ML gives a good prediction of the HTTR control rod position with less than 5% difference compared to that in the experiment. With increasingly complicated experiments that create a large amount of data, ML is also expected to improve the design and safety analysis of the HTTR in the future.
Analysis of Energy Dissipation Effects of Resistance Components in the Water Hammer Process
Authors: Yang Linqing , Qin Benke , Bo Hanliang
Published: November 2022

Water hammer is common in nuclear engineering pipeline systems, such as feedwater systems, essential service water systems and passive residual heat removal systems, etc. Water hammer can cause a large fluctuation of fluid pressure, posing a threat to the normal operation of pipeline systems, so it is necessary to reduce the pressure fluctuation and energy. The flow resistance components such as the orifice plate are the key parts to mitigate the influence of water hammer. For the reservoir-pipe-valve (RPV) system, a water hammer model that considered the effect of unsteady friction was established and solved by the method of characteristics. The calculation results of theoretical model agreed well with the experimental results. On this basis, the orifice plate was added to analyze its influence on the transient flow process. Based on the traditional analysis method and energy analysis method, the dissipation effects of the frictional resistance and the local resistance of pipeline system were studied. The analysis results show that as the orifice plate moves from upstream to downstream of pipeline system, the fluctuation of internal energy and kinetic energy decreases, the orifice dissipation work increases, and the friction dissipation work decreases. The research results show the advantage of energy analysis method and provide guidance for the design of energy dissipation equipment.
Countercurrent Flow Limitation Calculation of Hot Leg Based On System Program
Authors: Li Xiang , Sun Wan , Wen Zhu , Zhang Luteng , Ma Zaiyong , Pan Liangming
Published: November 2022

In the event of a Loss of Coolant Accident, a large amount of coolant is released from the reactor, and steam is generated inside the reactor, and then flows down the heat pipe into the steam generator. When a natural circulation is not established in a loop, the condensed liquid in the steam generator is returned to the core by gravity of the wall of the pipe through the hot leg. This countercurrent flow has important influence on reactor safety analysis. At present, many researches have been carried out in this field, but most of them focus on the study of the countercurrent flow characteristics of two-phase flow in vertical or horizontal pipes, there are few researches on the countercurrent flow in a complex. In this paper, the countercurrent flow of two-phase flow in the hot leg under the condition of Loss of Coolant Accident (LOCA) is modeled and calculated. The experimental loop is composed of vertical pipe, inclined pipe and horizontal pipe. It is found that the shape of the hot leg has a significant effect on the injection limit, and the data based on dimensionless square root of velocity is compared with Wallis’ vertical tube results. The effects of liquid viscosity, density of liquid phase and gas, and geometry of pipeline on flooding were also studied.
Simulation of Detection Efficiency of High Integrity Container Gamma Scanning Device for Nuclear Power Plant Based on CZT Detector
Authors: Xie Wenzhang , Huang Jia , Shan Chenyu , Lei Qingxin , Chang Kang , Zhou Hao , Liu Feng , Lin Peng , Liu Xiajie , Li Li
Published: November 2022

High Integrity Container (HIC) for nuclear power plant is an advanced waste reconditioning Container specially designed and manufactured. It has the characteristics of High strength, good sealing, strong chemical stability and thermal stability. It can be used to load a variety of Low and Intermediate Level radioactive wastes that have not been cured or fixed, such as dehydrated mud and evaporation residues, Waste resins and Waste water filtration cores. According to relevant national standards and nuclear power plant radioactive waste management regulations, it is necessary to measure the type of radionuclide and radioactive level of waste barrels. The sizes of HIC barrels are DI, DII and DIII, with volumes ranging from 0.42 m3 to 3.44 m3. There are various types of waste in the bucket, and the waste density is not less than 0.25 g·cm−3; The main material of HIC barrel is high density crosslinked polyethylene, which has poor shielding effect on γ-rays, resulting in high dose rate level on the surface of the barrel. Therefore, the existing γ scanning device for 200 L/400 L metal waste drums in nuclear power plants cannot meet the measurement requirements of HIC drums, and it is necessary to develop a special γ scanning device for HIC drums. In this study, Cadmium Zinc Telluride (CZT) semiconductor was used as the main detector to simulate the detection efficiency under different HIC barrel sizes, different waste density, different detection distances and different collimating opening diameters. The simulation results show that the CZT detector can be used for gamma scanning measurement of HIC drums, especially HIC drums with high dose rate, under the condition of proper shielding collimation and detection distance.
Three-Dimensional Numerical Simulation of the Release and Migration Behavior of Fission Gas Bubbles in Advanced Reactor Pool Under Serious Accident
Authors: Mai Zijun , Zhao Zichen , Cheng Hui , Cheng Songbai
Published: November 2022

Three-dimensional numerical study on the release and migration of bubbles in liquid metal pool is carried out for severe accidents in advanced reactors. The fission gas bubble migration will affect the release and distribution of radioactive materials in the molten pool, and then affect the evaluation of the source term in the reactor. The simulations are conducted based on the Volume of Fluid (VOF) method for both water and molten Lead Bismuth Eutectic alloy (LBE). The bubbles are initialized with different diameters or generated by gas injection. The methodology is firstly validated by the simulations of single bubbles rising in water. The simulation results, such as the rising velocity and aspect ratio, of the bubbles with different equivalent diameters agree well with the experiments. Then the model is applied to predict the release and migration behavior of bubbles in molten LBE. It is found that the terminal rising velocity and aspect ratio of single bubbles in LBE have similar profiles to those in water. The rising velocity increases rapidly at first, and then fluctuates within a certain range. The bubbles in molten LBE have similar behavior to typical bubble migration in liquid. At the typical gas injection speed of 1 m/s, the bubble coalescence is observed in molten LBE. The distribution of the terminal rising velocity and equivalent diameter has a bifurcation which is influenced by the bubble wake. The results of the simulations with molten LBE are comparable to theoretical results. This work provides better understanding of the fission gas bubble migration behavior under severe accidents in advanced reactors and will contribute to the safety analysis as well as the design optimization for lead-cooled fast reactors.
Effect of Temperature and Potential on BCC-Fe Nanoindentation Process
Authors: Wang Dexin , Lin Pandong , Sun Libin , Cai Zhipeng
Published: November 2022

The irradiation mechanism of iron-based alloys has proven to be related to dislocation movement and interaction between dislocation and irradiation defect. Fe is an important element of the iron-based alloys. The single crystal iron is selected as the material of the indenter and the substrate when molecular dynamics (MD) simulation is launched. This article studies the effect of the different temperature and potential functions on the dislocation evolution and mechanical properties of single crystal iron in the nanoindentation process.

Results show that temperature affects the load-displacement curve; the potential has no effect on the trend of the load-displacement curve, but it has an impact on the value of the load; the dislocation behavior of the substrate after compression results in a jagged load-displacement curve (pop-in event). These results are of great significance for understanding the dislocation behavior of the typical body centered cubic (BCC) metal during nanoindentation process.
Summary of Effects of Oxide Layer on Permeability Coefficient of Tritium
Authors: Zhou Ziling , Wang Yu , Guo Jingni , Xie Feng , Li Wenqian , Wen Yanwei , Shan Bin
Published: November 2022

Tritium is an important nuclear fuel in fusion reactors and one of the dominant nuclides in the primary coolant in fission reactors. The storage, feeding, control, monitoring, and transport of tritium are important for practical engineering applications. Because of the high mobility of tritium in both fission and fusion nuclear systems and its effect on human body, tritium has received great attention worldwide. The retention and prevention of tritium permeation from primary loop to secondary loop is a common research interest in various reactors. Previous studies indicated that a surface oxide layer is an efficient method to reduce tritium permeation. In this paper, we summarize the effects of various oxide layers on the permeation of hydrogen isotopes in nuclear reactors. The permeation reduction factor for materials used in fusion reactors ranges from 1000 to 100000. The diffusion behavior of tritium in several materials with and without oxide layer is discussed in detail. The oxide layer is more important than intrinsic permeability to prevent tritium permeation. As tritium is the only radioactive source term in the secondary loop of high temperature gas-cooled reactors, we also reviewed the permeation of hydrogen isotopes in the heat exchanger, which is an important issue in nuclear hydrogen production. The present study provides a comprehensive overview of tritium permeation behavior and is expected to promote the development and design of tritium-permeation-proof materials in the future.
Design of the Protection Device Withstand External Explosion Shock Wave for the Nuclear Power Plant
Authors: Zhang Mingqian , Li Zhenguang , Wang Qingli
Published: November 2022

The design of the nuclear power plant shall meet the general objective to ensure that the safety functions of the systems and components are not affected by any external explosion hazard. A protection device must be sufficiently designed to protect the target equipment from the external explosion load. The shock wave propagation and attenuation was simulated dynamically for a new protection device withstand external explosion by use of the computational fluid dynamics software FLUENT. The method used in the simulation was firstly validated by comparing the simulation and the results of the shock tube test. Furthermore, the numerical model of the shock wave propagation and attenuation through the protection device was developed by this method, and the pressure time history behind the protection device was obtained, and then the protective effect of more layers of the protection device and the attenuation effect of the protection device preventing the lower amplitude of the incident shock wave were taken into account. It was found that the protection device could provide an effective limitation to the external explosion, the protective effect had not been strengthened obviously with more layers of the same protection device, and the attenuation effect would decrease with the lower amplitude of the incident shock wave. This analysis practice predicts virtually the physical phenomena, and can be taken as a guidance for the protection device design development and improve our understanding of external explosion shock wave phenomena.
Control Strategies Evolution of Nuclear Plant: From Obninsk to HTR-PM
Authors: Cheng Zhonghua , Dong Zhe , Li Bowen
Published: November 2022

Since the first commercial nuclear plant Obninsk began to operate, there have been dozens of types of nuclear plants studied, designed, or constructed nowadays driven by the key issue: how to enhance the economic competitiveness of nuclear plants while ensuring safety. Research on the key issue stimulated the evolution of nuclear plants, which simultaneously brought an array of various control problems for each generation and ultimately attributed to the evolution of control strategies of nuclear plants. In this paper, the main development route of nuclear plants is presented with a concise introduction to the representative design of nuclear plants for each generation. Meanwhile, the corresponding control problems brought by the design and/or the safety requirement document, national and international requirements, and so forth, were analyzed with emphasis on the Gen IV multi-modular high-temperature gas-cooled reactor (MHTGR). On account of the requirement to cooperate with other renewable energies to achieve the “Emission peak, Carbon neutrality” through the improvement of operation flexibility of NPP, most SMR plants work in the reactor following mode which requires more R&D efforts on the coordination control and load-following control to enhance economic advantages of SMR plants, etc.
Development and Verification of Containment Heat Sink Module in Severe Accident Code
Authors: Yang Zhi , Zhou Peng , Chen Lin , Cheng Yixuan , Zhang Hao , Yang Yanhua , Mao Yichen
Published: November 2022

As an important part of COSINE software package, cosSA code is used for safety analysis of severe accidents in nuclear power plants. The containment heat sink can transfer the heat from the containment to the environment and thus decrease the risk of the containment failure, especially when the primary loop is broken. The containment heat sink module provides heat transfer calculation from gas space and water pool to the heat sink, so as to accurately predict the temperature of containment heat structure and gas temperature and pressure in the compartments. In this paper, the development of the containment heat sink module in cosSA code is introduced, including lumped parameter heat sink, distributed parameter heat sink, heat sink surface heat transfer, and inner heat conduction. In order to verify the current containment heat sink module, a series of tests are designed based on a simple case with three compartments. In general, the test results are in good agreement with the similar software for the totally submerged and totally uncovered heat sink, indicating the effectiveness and accuracy of the current containment heat sink module. For the partial submerged distributed parameter heat sink, the transfer to the gas space and water pool are calculated simultaneously. The axial division method can alleviate the extra heat transfer effect from gas space to water pool.
Development and Verification of Multigroup Cross Section Generation Module GroupXS in AXSP
Authors: Hu Kui , Ma Xubo , Ma Xuan , Zhang Chen , Jia Guanqun , Chen Yixue
Published: November 2022

It is important for nuclear reactor physics calculation to generate accurate multigroup cross section libraries by using evaluation nuclear data libraries. A new evaluated nuclear data library processing code, named as AXSP (Advanced Cross Section Processing Code) was developed with independent intellectual property rights. The GroupXS module was developed with the ability of generating multi-group cross-sections based on the Bondarenko method which is always a favorable approximation for the fast spectrum reactor, and the method was also used in GROUPR of NJOY. To verify the accuracy of the GroupXS module, the GROUPR module of NJOY was used, and we corrected three numerical verification bugs of GROUPR. The bugs were described in this study. By comparing the calculation results with GROUPR of NJOY2016, the relative error of multigroup flux, multigroup cross sections of various reaction types are less than 0.01%, and the relative errors of transfer matrices of various reaction types are basically less than 0.1%. All the neutron reaction types of all nuclides in ENDF/B-VII.1, ENDF/B-VIII.0 and CENDL-3.2 can be processed by GroupXS, and the multigroup cross sections generated by GroupXS of AXSP have a good agreement with that generated by GROUPR of NJOY2016.
A Space Reactor Control System Design and Simulation Platform Development and Verification
Authors: Liu Pinchao , Sun Peiwei
Published: November 2022

A liquid metal cooled reactor coupled by mixed gas of helium and xenon closed Brayton cycle is designed for space exploration purpose. The design and dynamics of the space reactor differ from those of traditional reactors. Therefore, it is necessary to develop a control system design and simulation platform to investigate its dynamic behavior, design and evaluate the control system. Two system codes are adopted and coupled to develop the platform. The reactor core system code is used to calculate the reactor kinetics and thermal-hydraulics of the coolant. The energy conversion system code is to compute the Brayton cycle and the heat exchanger between the primary and secondary loops. 3KeyMaster simulation platform is used to load the two system codes by an interface program through dynamic link library. The data exchange and time synchronization between 3KeyMaster simulation platform and two system code are performed to realize the coupling of two system code. By introducing typical disturbances, the calculation results are compared with those from the system codes to verify the interface program. The input and output parameters can be visualized by model components in 3KeyMaster simulation platform to analyze the dynamic behaviors, which is useful for the subsequent control system design.
The Location of Spent Fuel Pins in Heavy Water Research Reactor in the Spent Fuel Pool
Authors: Lu Jin , Nie Peng , Ren Ren , Guo Yaping , Wang Qi , Zhang Xinwang , Zhang Lijun
Published: November 2022

Heavy Water Research Reactor was the first reactor in China. It adopted metallic uranium as fuel and aluminum alloy as cladding in the beginning. During the refueling procedure many years ago, several fuel pins dropped at the bottom of the spent fuel pool. The amount and the location of the fuel pins were not recorded. Before decommissioning, it is important to locate and take back the fuel pins. The fuel pin is φ40 mm × 100 mm and the pool is 17.8 m × 5.6 m. It is like dredging for a needle in the sea. What’s worse, it is not clear to find the fuel pins even with a waterproof camera for the bad water quality. Besides, radiation tubes, neutron detectors, and broken control rod were also stored in the pool.After reviewing on the localization method of radioactive source at home and abroad, it is determined to use γ dose rate meter and γ spectrometer in this case. Shields in different width and thickness were calculated with MNCP code. The optimization was 5cm in thickness and 5 cm width lead shield with a hole in the center. The waterproof camera was tied with Gama detector and finally 14 fuel pins were located and taken back safely.
Simulation Research on Real-Time Offsite Dose Radiation of Nuclear Power Plant
Authors: Ou Pingwen , Wang Zhenghui , He Dongyu , Jiang Pingting
Published: November 2022

With the rapid development of nuclear power, its safety has attracted more and more attention. In the event of a nuclear accident, once the leakage of radioactive material occurs, it will pose a huge threat to the surrounding environment and the economy and safety of human. In this case, the rapid and accurate evaluation of pollutant dispersion and radiation dose of offsite consequences can guide the decision of nuclear emergency, including the public emergency response and rescue operations. It can also provide the best solution of routes for public evacuation and rescue entrance. The limitation of radioactive accident consequences is essential to protect property security and public safety.

In this paper, a real-time evaluation program for offsite dose for nuclear emergency system is developed. The selected for atmospheric dispersion model is the segmented Gaussian model (SGM). Diffusion, advection, plume rise, dry deposition, wet deposition, decay, etc. are considered and FORTRAN language is used for the development.

For code-to-code comparison, C3X, the French general offsite dose evaluation program, is chosen as the reference. In the release of changing weather conditions, the simulation results are consistent with C3X, which demonstrates the validity and reliability of program. This paper makes also a nuclear emergency simulation for the changing weather conditions. The results show that the program can precisely calculate the plume nuclides concentration, ground concentration and dose in different time and space. It can also accurately divide the dose area for public emergency guide and rescue operations.
Modal Simulation and Verification on Stator End Winding of Large Nuclear Power Turbine Generator
Authors: Yang Xike , Zhang Jian
Published: November 2022

The 1300 MW nuclear power turbine generator is a four-pole generator with a rated speed of 1500 r/min. Due to its large capacity and low speed, the stator core and stator end winding bear huge electromagnetic torque. At the same time, because of the large heat generated by the stator winding during operation, the stator end winding is subjected to a large thermal expansion force. All these loads may have a serious impact on the stator end winding. In this paper, the three dimensional parametric finite element method and modal optimization algorithm are used to obtain the finite element model of each component of the stator end winding. Finally, the modal natural frequency of the overall end winding structure is obtained, so that the four-lobe modal natural frequency avoids the excitation frequency of 100 Hz, and the accuracy of the simulation mode is verified by modal test. The stator end winding modal simulation provides a reference for the stator end winding structure design of the large nuclear power turbine generator, ensures the good dynamic characteristics of the end winding structure, which provides a guarantee for the safe and stable operation of the stator end winding.
Modelica Based Economic Analysis of Nuclear Fuel Cycle
Authors: Liu Jie , Wang Xiang
Published: November 2022

The nuclear fuel cycle with spent fuel reprocessing process is called closed cycle. The reprocessing process of the fuel cycle recovers the fission elements uranium, plutonium and their isotopes. The cost of the nuclear fuel cycle includes front-end and back-end. The front-end segments mainly include mining, conversion, concentration, fuel element fabrication, etc. The back-end cost includes storage, processing, disposal and so on. Modelica language is a physical process description language based on objects and equations. Modelica language can be used to realize the visual modeling of various causal or non-causal processes. The main content of this research is to model different nuclear fuel cycles, including summarizing the simplified model of closed fuel cycle system, economic considerations and system working principle, the current economic analysis and research status of closed fuel cycle, combing the existing models, the use and application of OpenModelica language, especially the visual modeling method of OpenModelica, to achieve economic analysis of different nuclear fuel cycles.
A Weakly Supervised Time Series Analysis Framework for Anomaly Detection in Nuclear Power Plants
Authors: Dong Feiyan , Chen Shi , Demachi Kazuyuki , Yoshikawa Masanori , Seki Akiyuki , Takaya Shigeru
Published: November 2022

Condition monitoring is essential to the management and maintenance of Nuclear Power Plants (NPPs), as anomalies in the condition of components can affect the normal operation state of the entire plant. Therefore, timely and automatic detection of anomalies plays an important role and is in high demand. At present, deep learning is widely used for anomaly detection. Nevertheless, anomalies are difficult to define, sparsely occurring, and are accompanied by variable noise labels, which poses challenges to detection. Moreover, the problems such as loss of temporal features and gradient vanishing that exist in general deep learning models when dealing with time series data also increase the difficulty of anomaly detection. In response to these problems, a weakly supervised time series analysis framework for anomaly detection in NPPs is proposed, constituted of weakly supervised learning (WSL) and attention mechanism. The validation of the proposed framework was performed on the High Temperature Engineering Test Reactor (HTTR) anomaly cases dataset using the analytical code “ACCORD”, which distributed anomalies independently across multiple instruments and were recorded from the responding sensors to each anomaly. At this stage, 3 classes of anomalies were used as input data for the validation experiments. The experimental results demonstrate the effectiveness and feasibility of the proposed framework on anomaly tasks for the condition monitoring of NPPs.
Developing Intelligent Equipment Monitoring System of Nuclear Power Plants to Improve Plant Performance
Authors: Yu Jianming , Cai Yibiao
Published: November 2022

The active development of nuclear power is an important strategic choice to achieve carbon peak and neutrality goals. It is an essential study to realize the economic operation of nuclear plants on the premise of ensuring the safety and reliability of nuclear power plants. Intelligent monitoring and maintenance technology is an important development tendency that can effectively improve the operation safety and maintenance economy of nuclear power plant. In this article, an intelligent monitoring and maintenance system of AP1000 significant equipment in Sanmen Nuclear Power Plant is developed based on the equipment operation mechanism and failure mode analysis by using advanced pattern recognition technologies including intelligent sensing, big data analysis, multidimensional spatio-temporal correlation condition monitoring and fault diagnosis technologies. It can realize online real-time equipment performance monitoring, automatic trend analysis, automatic health condition evaluation, early anomaly identification and automatic fault diagnosis of equipment in nuclear power plant and promote the transformation of maintenance mode from time-based maintenance to condition-based maintenance. The intelligent monitoring and maintenance system has been applied in various mechanical and electrical equipment of Sanmen Nuclear Power Plant such as reactor coolant pump, main steam turbine, main transformer, heat exchanger and achieved certain technical achievements and economic benefits. The construction of intelligent monitoring system has certain demonstration and popularization.
Estimation of Decommissioning Radioactive Wastes Quantity for Nuclear Power Plant
Authors: Li Senquan
Published: November 2022

Most of the radioactive waste generated during the decommissioning of nuclear power plants are relatively low in radioactivity, except for the reactor pressure vessel, reactor internals and primary shielding walls, which have high levels of radioactivity. However, the treatment and disposal costs of decommissioning waste account for a high proportion of the total decommissioning cost. In order to minimize the decommissioning waste and to plan the decommissioning costs, it is necessary to study and estimate the total quantity of decommissioning radioactive waste in the design phase. This paper presents the statistical method for estimation of decommissioning radioactive waste quantity, gives the complete steps for estimating decommissioning radioactive waste quantity, and determines the waste quantity estimation principle for each type of decommissioning radioactive waste. Finally, the estimation task based on a Chinese typical million kilowatt nuclear power station single unit is reasonably completed. It is estimated that the quantity of decommissioning radioactive waste generated by this nuclear power station single unit is about 12000m3, which is similar to that of the same type-unit in the world. This proves the rationality of the method for estimating decommissioning radioactive waste quantity proposed in this paper.
Experimental Study on Measurement Characteristics of Different Type Flow Meters in Pb-Bi Environment
Authors: Liu Xin , Yun Xiangyu , Liang Junming , Zhang HuiYong
Published: November 2022

Lead bismuth eutectic is one of the most promising materials for coolant of fourth-generation reactor due to its outstanding heat transfer performance and chemical stability. However, the density and viscosity of lead bismuth eutectic is quite different from water, leading to a difficulty for flow measurement because of the high temperature operating environment and its strongly corrosive. Therefore, there is currently not a reliable flowmeter for measuring the volume flow of the lead bismuth eutectic. Electromagnetic flow meter has the advantage for measuring lead bismuth eutectic flow, due to the non-contact measurement, simple structure and high temperature resistance, so it has become one of the main selection flowmeters for lead bismuth eutectic flow measurement. And venturi flowmeter is one of the most widely used flowmeter types in water media. Therefore, the measurement characteristics of the electromagnetic flow meter and the venturi flowmeter in liquid lead bismuth eutectic environment had been studied.

In this context, two type flow meters had been calibrated by the volume method. The influence of the temperature and other factors had been studied on these two flowmeters.

The experimental results showed temperature affected the electromagnetic flowmeter measurement results, but it had no effect on venturi flowmeter. And the flow coefficient of electromagnetic flowmeter has a linear relationship with the temperature. And the relationship between temperature and flow coefficient of electromagnetic flowmeter was fitted by the calibrated data. Also, the flow coefficient of venturi flowmeter had been corrected too.

Also, the influence of discharging and refilling the LBE had been studied. And the results showed there is no influence on two type flowmeters. So, there is no necessary to calibrate the flowmeter repeatedly and the experimental process of LBE can simplify.