bluuebunny/crossref_metadata_2025_split huggingface dataset:
The Dose Assessment Method Based on the Local Method of Characteristics and Skeletal Animation
Authors: Cao Lun-xiu , Chao Nan , Liu Yong-kuo , Chen Zhi-tao 
Published: November 2022

Abstract
Flexible and accurate estimation of dose absorbed by critical human organs is important to ensure and improve workers’ radiation safety. Absorbed doses for different organs and tissues are affected by their positions and the orientation of the body in the radiation field. This paper presents a skeletal animated local method of characteristic (SALMOC) dose assessment method that considers the working dynamics and organ-level dose for occupational workers during nuclear facility decommissioning. The proposed method combines the verisimilitude of skeletal animation technology (SA) with the flexibility of the local method of characteristics (LMOC). To account for the working posture, the change in different working postures are controlled by skeletal animation, and then the voxel model in a certain pose is generated according to the changed human model. Finally, the LMOC is used to perform the dose assessment. The proposed SALMOC method not only accounts for different working postures during the decommissioning of nuclear facilities but also considers the occlusion and scattering effect of the human model to achieve accurate organ-level dose assessment. The effectiveness of the proposed method is verified by comparing the results with those derived from the Monte-Carlo method and Point-Kernel method evaluated on two different organs. The result shows that the proposed method can generate voxel models for irregular organs, and the tests show better consistency with the Monte-Carlo method compared with the Point-Kernel method.
Application of Probabilistic Safety Assessment in the Risk Analysis and Management of Interface Systems LOCA in a Nuclear Power Plant
Authors: Liu Leilei , Liu Weidong , Wang Zhao 
Published: November 2022

Abstract
Interfacing systems LOCA is a kind of special loss of coolant accident in nuclear power plants, which has an important impact on the nuclear power plants safety. Because the break of this kind of accident is located outside the containment, the interfacing systems LOCA will cause a large amount of radioactivity to be directly released from the core to the atmospheric environment without the detention of the containment, so which is an important contribution to the large early release of nuclear power plants. The Probabilistic safety assessment have been applied to a nuclear power plant in Chinese mainland in this paper. The risk of interfacing systems LOCA in the nuclear power plant is comprehensively evaluated from the occurrence scenarios and frequency of initial events, accident mitigation strategy and event tree modeling, accident consequence and impact analysis. Through the application of probabilistic safety assessment, the key factors that may lead to a core damage and large release of the interface LOCA in the nuclear power plant has been find out, and the weakness existing in the power plant in dealing with such accidents has been discussed at the same time. According to the results of the analysis, how can the nuclear power plant deal with such accidents much more better is discussed, the improve suggestions are given.
Investigation of Swelling Behaviors of U-10Zr Metallic Fuel in the Low Temperature Regime via a Combined Thermo-Mechanical and Rate Theory Analysis
Authors: Qian Zhengyu , Yun Di 
Published: November 2022

Abstract
Uranium zirconium (U-Zr) metallic nuclear fuel has long been considered as a promising candidate for fast reactor fuel in the past few decades duo to its excellent performance. One of the key performance problems of U-10Zr metallic fuel is its strong swelling which would lead to significant fuel cladding mechanical interaction (FCMI) at high burn-ups. Therefore, the mechanism of fuel swelling has long been a concern of great interest in a wide range of this fields. Recently published research conducted by the Institute of Nuclear Physics and Chemistry in China got the experimental results of U-10Zr fuel swelling in a low temperature regime (400–600K) which showed perceivable swelling of the in-pile irradiated fuel. The previous exploration on the swelling mechanism of uranium zirconium metal fuel at low temperature can’t be carried out due to the lack of data and this research filled the gap for the swelling data of uranium zirconium metallic fuel at low temperature. In our work, combined with the input parameters given by finite element analysis, we utilized a rate theory code based on the cavitational void swelling model to catch the fuel swelling behaviors of the U-10Zr fuel in the low temperature regime building upon these data. And a new set of key parameters for the rate theory model were also determined.
Manufacturing Technology of Thick-Walled Integrated Cylinder With Large-Sized Side Nozzles for Nuclear Pressure Vessel
Authors: Li Shaofei 
Published: November 2022

Abstract
This paper introduces a manufacturing technology of thick-walled integrated cylinder with large-sized side nozzles for nuclear reactor pressure vessel. This forging integrates the inlet nozzle and the outlet nozzle on the nozzle shell with integral flange to form an integrated forging. China First Heavy Industry (CFHI) has successfully realized the manufacturing process of this forging. During the forming stage of the forging, with the help of numerical simulation method, the size of the forging blank is reduced to the greatest extent, and a thick-walled cylinder with a convex structure is designed and forged, and then the punching guide hole is machined at the convex structure of the cylinder, which plays the role of positioning and guiding in the forging process of the side nozzle. We use special-structured forging aids to transfer the load of the forging press to the punch to complete the hot extrusion of the side nozzles, and finally realize the integrated near-net shape of the forging. The forging flow line of the forging is preserved to the greatest extent, and the best forging effect of each part of the forging is guaranteed. After the final heat treatment process of the forging, all parts of the forging obtained excellent mechanical properties.
Research and Development of SA508Gr.4N Steel Ultra-Thick Forgings With High Strength and Toughness for Nuclear Engineering
Authors: Du Dongxu , Zhao Deli , Ren Liguo , Zhang Wenhui , Li Jiaju , He Xikou 
Published: November 2022

Abstract
Due to low hardenability, the conventional SA508Gr.3 steel cannot meet the design requirements of increasing wall thickness of forgings for nuclear power pressure equipment. In this paper, the hot processing and manufacturing technology of SA508Gr.4N steel ultra-thick forgings for the new generation of nuclear reactor pressure vessel was introduced. The engineering development of nuclear SA508Gr.4N steel forgings with wall thickness from about 700mm to 1000mm was realized by China First Heavy Industry (CFHI). The largest forging was made of 228 tons ingot and the final maximum dimensions reached about ϕ4000mm × 1000mm. The samples were taken anatomically from the forging at various thickness and radial positions, and subjected to the tensile, Charpy impact and Drop-weight tests. As a result, the full-section property of the forging met the requirements of the ASME code. Compared with nuclear SA508Gr.3 steel forgings, the strength and toughness of SA508Gr.4N forgings were greatly improved. Meanwhile, based on the fact that the actual property value is much higher than the requirements of ASME code, higher design property requirements of SA508Gr.4N steel forging was proposed. In addition, the nuclear neutron-irradiation-embrittlement data of SA508Gr.4N steel, which were cut from the actual engineering forgings and experimented in a high flux engineering test reactor (HFETR), was obtained. The main research and development goals for engineering application of SA508Gr.4N steel in the future were also discussed.
Finite Element Analysis and Prediction of the Cross-Section Morphology of Grid Strips
Authors: Mao Jianzhong , Cheng Zhi , Su Jun , Wang Ben , Zhang Pengfei , Mei Zezhi , Yuan Jiajian , Gao Bo 
Published: November 2022

Abstract
To investigate the relationships between the quality of the section morphology of the zirconium alloy strips and the blanking parameters as well as the design of the blanking mould, 64 processing parameter combinations are generated by using Latin Hypercube Sampling and the finite element models with different boundary conditions have been constructed to acquire the corresponding section qualities. The calculation results of the finite element model are verified by the experiment data. Furthermore, the random forest algorithm is employed to build the prediction model between the processing parameters and the lengths of the section bright strip. The sensitive analysis of the processing parameters influenced on the section quality have been conducted through the prediction model. The results indicate that the random forest prediction model have a satisfied accuracy, which could effectively forecast the length of the bright band in the section, and then provides guidance on mold design as well as process parameter selections. The blanking clearance included in the design factor has a decisive influence on the length of the bright strip; The friction coefficient has the weakest effect on the results. The effect of die fillet radius and blanking speed on the length of bright band is not significant.
Measurement of the Two-Phase Flow Void Fraction Downstream Of Spacer Grids in Tight Lattice Bundles Using Wire-Mesh Sensor
Authors: Yan Xu , Xiao Yao , Zhang Hengwei 
Published: November 2022

Abstract
The tight lattice fuel assemblies (P/D &lt; 1.1) have huge application potential in the new generation of the nuclear reactor systems, which possesses the advantage of higher power density and higher fuel conversion ratios, etc. Spacer grids can fix the fuel rod position and enhance the heat transfer performance. However, the two-phase flow characteristics downstream of spacer grids in the tight lattice are still not clarified. In this paper, a gas-liquid two-phase flow experiment study measuring the void fraction distribution downstream of spacer grids without mixing vanes (SGWMVs) in a double subchannels tight lattice bundle was conducted using a double-layer wire-mesh sensor. The experiment channel and the SGWMVs were both up-scaled (1:2.7) considering the difference between the experimental flow conditions and the prototype flow conditions. The phase distributions for a series of flow conditions (0.07 m/s &lt; jg &lt; 1.04 m/s, 0.93 m/s &lt; jl &lt; 1.86 m/s) were measured at Z/Dh = 115.81 of the bare rod channel as the flow data upstream of the spacer grid. Whereafter, the SGWMVs was installed at the position of Z/Dh = 145.92. The phase distributions 45mm downstream of SGWMVs were measured successively for the same flow conditions. Three types of flow patterns were obtained, including the bubbly flow, the cap-bubbly flow, and the slug flow. And the effects of SGWMVs on different flow patterns were described and analyzed. The newly obtained data established the reliability database contributing to the development of the computational fluid dynamics codes and the interfacial area transport equation.
A New Correlation for Predictions of Heat Transfer of Lead-Bismuth Eutectic in Circular Tubes
Authors: Wang Zhengquan , Wen Qinglong , Xu Shijia 
Published: November 2022

Abstract
Liquid lead or lead-bismuth eutectic has good thermal physical properties, neutron performances, and natural circulation capacity. It is the first choice of coolant material for the lead-based fast reactor (LFR). However, there is a fundamental problem, heat-transfer characteristics and quantitative relationship of liquid lead and lead-bismuth eutectic, to be solved in the process of thermal-hydraulic design and safety analysis code development of lead-based fast reactor. The existing correlations are mostly fitted by the experimental data under specific conditions and the applicability is poor in a wide range of parameters, which affects the accuracy of thermal-hydraulic design and safety analysis results. Therefore, it is a necessity to establish experimental databases and correlation databases to obtain a more accurate heat transfer correlation of liquid lead and bismuth eutectics within a wider range of parameters.
This paper reviews the typical theoretical and experimental studies on the heat transfer characteristics of liquid lead and lead-bismuth eutectics in the recent 70 years, and an experimental database based on the heat transfer experimental data in the published literature is established, which data points include a wide range of thermodynamic parameters — inner diameter ranging from 9.00 to 16.56 mm, temperature from 250.0 to 412.80 °C, heat fluxes from 20.0 to 12900.0 kW/m2, and mass fluxes from 0 to 12.0 kg/m2s. Based on the database, the typical heat transfer correlations of liquid lead and lead-bismuth eutectics are evaluated and analyzed, and a new heat transfer correlation is developed, its suitable range is 100 &lt; Pe &lt; 5800. The results show that the mean absolute relative error of the new correlation is 16.64%, the average relative error is 0.74%, the standard deviation is 4.37%, and the error of 65.27% of the total data points is within ± 20%. Therefore, the heat-transfer correlation proposed in this study can well predict the heat-transfer coefficient of high-temperature liquid lead or lead-bismuth eutectic in the tube.
Thermal-Hydraulic Safety Analysis of Natural Circulation Lead-Cooled Fast Reactor SNCLFR-100 Core Based on Porous Medium Approach
Authors: Feng Wenpei , Yang Guangliang , Zhang Kefan , Qin Chong , Luo Xiao , Ding Tao , Chen Hongli 
Published: November 2022

Abstract
In this study, a thermal-hydraulic analysis is carried out for the core of natural circulation lead-cooled fast reactor SNCLFR-100. Steady-state and transient analysis are performed with porous medium approach-based code TWOPORFLOW. In the steady-state analysis, mass flow distribution and temperature distributions of the assemblies are analyzed in assembly-wise mode. At the same time, the hottest assembly is analyzed in pin level and the safety performance is investigated. In the transient analysis, a typical Design Extension Conditions unprotected over-power transient is simulated with one-way coupling method.
Investigation of Multifarious Zircaloy-Water Oxidation Models on Core Degadation in a Typical SBO Accident
Authors: Sun Hongping , Luo Yuejian , Deng Jian , Zhang Ming , Wu Xiaoli , Liu Lili , Xu Youyou 
Published: November 2022

Abstract
Cladding oxidation is a particularly important process which can affect the process of core melting. Synchronously, hydrogen released from zircaloy-water reaction is a threat to the integrity of containment. Up to now, multifarious zircaloy-water oxidation models are applied in different severe accident analysis programs. In order to investigate the influence of different cladding oxidation models on the core degradation process, a station blackout (SBO) accident model of CPR1000 nuclear power plant (NPP) was established and simulated by modifying the cladding oxidation model in this work. Five oxidation relations were chosen for evaluating thermal hydraulic, metal oxidation and core degradation analysis. Thermal hydraulic characteristics of five types remain consistent before core oxidation. For more concerned hydrogen production, the results show that the ratio of the maximum hydrogen production to the minimum is approximately 2.5 times in comparison of the different oxidation models. Four core nodes temperature distributions at different time points are selected to analyze the core node degradation process, which can help readers understand the effect of different models on core degradation more intuitively. The oxidation heat generated is consistent with that of hydrogen production, indicating that the core degradation time in the case of less hydrogen produced is delayed than that of more hydrogen produced. It follows that different oxidation models have great influence on the core oxidation behavior, which is of great significance to understand different oxidation models of zircaloy and the core degradation behavior of severe accident. It is critical to choose the appropriate model to simulate the whole severe accident.
Analysis of Water Consumption in Nuclear Power Plants and National Standard Formulation for Norm of Water Intake
Authors: Wang Rui , Xu Kechen , Zeng Kehao , Yan Hongmei , Bai Wei 
Published: November 2022

Abstract
The nuclear power industry consumes a large amount of water. Its water process systems are complex, and require high water guarantee rate. In general, water process and water consumption indexes for different nuclear power plants are still in the exploratory stage. For a long time, there are huge differences between the design values and actual values of water intake, water consumption, and water withdrawal in China nuclear power industry, which brings difficulties to the management. At present, the construction of nuclear power plants in China is in a period of rapid development, but there is no national standard for norm of water intake, yet. Nearly all the design water consumption indexes of nuclear power plants are based on the design values of nuclear power plants that have been built and part water consumption indexes required by few relevant design specifications. With the implementation of the strictest water resources management system, total volume control and norm management have become important tools of water resources management. The national standard formulation for norm of water intake in China’s nuclear power industry can provide a basis for the approval of water intake permits for design and construction of the nuclear power plants, to strengthen the management of water resource and improve the water utilization efficiency of the nuclear power plants. This paper introduces the current situation of water consumption in each nuclear power plant and the basis for national standard formulating for norm of water intake.
Development and Regulatory Reform on Radiation Safety Training for Nuclear Technology Utilization in China
Authors: Zhang Yang , Chen Fangqiang , Chen Baolong , Jia Wei , Li Zhaotong 
Published: November 2022

Abstract
Radiation safety training for the use of nuclear technology is an essential requirement of the regulations, and an important means to ensure the health of radiation practitioners and the safety of radiation environment. After more than ten years of development, the scale of the previous off-line training model basically met the needs of the industry, but some drawbacks also gradually appeared, such as lack of regulatory basis and supervision, high pass rate and poor training effect, increased corporate burden, etc. Since 2020, the National Nuclear Security Administration (NNSA) has officially implemented the reformation of radiation safety training. A unified radiation safety training and assessment platform has been established, covering the training and assessment needs of 11 nuclear technology utilization disciplines, and the training and assessment data are connected with the regulatory platform. The training assessment has changed from “accredited training institutions and paid off-line training and examination” to “free online self-learning and on-site online examination”.
This paper introduced the basic situation of the radiation safety training surveys, carried out multi-dimensional data analyses on the reformation of radiation safety training and assessment, discussed the existing issues in rules and the division of responsibilities between the companies and agencies. Some suggestions, such as adjusting the requirements of rules and regulations, increasing the coverage of nearby assessments, and adding the practical operation assessment, were proposed.
Specifics of Calculating Thermophysical Properties of CO2 and R134a in Critical Point Using NIST REFPROP
Authors: Xie Hanqing , Pioro Igor 
Published: November 2022

Abstract
Thermophysical properties of various fluids (liquids, vapours, and gases) are the fundamental knowledge for the application of these fluids. A computer program can be considered as a very useful tool if it is able to calculate various thermophysical properties of various fluids within a wide range of pressures and temperatures from lower ones and up to critical and supercritical ones. NIST REFPROP is such a program. For tens of years, it can be possibly considered as the best one in the world. However, it is not absolutely perfect. In the previous versions(s) of the NIST REFPROP 9.1 and lower, three basic thermophysical properties, specific heat, thermal conductivity, and volumetric expansivity, were subjected to very significant variations within the critical region. They had almost infinite peak values in the critical point, which was a theoretical approach.
In 2018, the latest version was released, Ver. 10.0 (https://www.nist.gov/srd/refprop). It was updated with new fluids, a wider range of pressures and temperatures, and with improved equations / correlations. The objective of this paper is to check if the deficiencies occurred in the previous version(s) have been fixed in NIST REFPROP Ver. 10.0. In this paper, various thermophysical properties of CO2 and R134a have been calculated by using NIST REFPROP Ver 10.0. The results around the critical point have been analyzed.
Inadvertent Reactor Pit Flooding Risk Research on the Generation III Nuclear Power Plant
Authors: Xiao Lingmei , Zhang Bing , Wang Ming , Wang Zhiwen 
Published: November 2022

Abstract
In-vessel Melt Retention (IVR) system is a typical design feature of the the generation III nuclear power plant. In case of severe accident, enough water is injected into the reactor pit to cool the outside of the Reactor Pressure Vessel (RPV) to maintain the integrity of the reactor pressure vessel, so that the core melt is retained in RPV. The design concept of IVR is very meaningful for maintaining the integrity of RPV and reducing the risk of large radioactivity release, but on the other hand, it may also introduce the possibility of inadvertent reactor pit flooding during normal operation, which may threaten the integrity of RPV. At home and abroad, the research on IVR mainly focuses on whether its safety function can be realized, but less attention is paid to its potential negative impact. This paper takes the China generation III reactor with the IVR design feature for the study, carry out the possible inadvertent reactor pit flooding path analysis and frequency analysis. Based on the risk study, put forward the concerns to prevent inadvertent reactor pit flooding, which will provide guide for the subsequent IVR design and optimization.