bluuebunny/crossref_metadata_2025_split huggingface dataset:
Establishment of Engineering Economics Standard System for Nuclear Power Plants in China
Authors: Yang Wen , Zhang Shaolin , Huang Jun , Wu Feifei 
Published: November 2022

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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.