bluuebunny/crossref_metadata_2025_split huggingface dataset:
Preliminary Development of a Coupling Environment Based on MOOSE and OpenFOAM and Its Application on Plate Fuel Modeling
Authors: Liu Hongquan , He Yanan , Wu Yingwei , Su Guanghui , Tian Wenxi , Qiu Suizheng 
Published: November 2022

Abstract
High-fidelity analysis of nuclear reactors is beneficial to improving the efficiency of nuclear reactor designs. To enhance modeling fidelity, multi-physics coupling simulations are typically employed. To this end, several platforms for code coupling are under active development, including MOOSE, SALOME, et al. In this paper, the loosely coupling interface between MOOSE and OpenFOAM was developed, aiming to extend the Computational Fluid Dynamics (CFD) capacity for MOOSE. Specifically, process control, data transfer, and mesh projection of the coupled simulation are realized by the “MultiApps” and “Transfers” module within MOOSE and the “externalCoupled” module within OpenFOAM, where text-based data transfer is used. The CFD capabilities of OpenFOAM and the finite element analysis capabilities of MOOSE are fully utilized in this coupling environment. Subsequently, a conjugated heat transfer problem for plate fuel was conducted to demonstrate the feasibility of the developed interface, where variables consisting of temperature, convective heat transfer coefficient, and heat flux at outer surfaces are transferred between MOOSE and OpenFOAM. The predicted results of the plate fuel seem reasonable and the validity of the novel coupling code for MOOSE and OpenFOAM is preliminary confirmed.
Study on Radioactive Inventory, Accident Source Term and Offsite Dose Consequence of DHR400 Under Fuel Handling Accident and Core Damage Accident
Authors: Wang Fenglong , Mei Xuanlan , Zhang Liang , Yang Yong , Zhang Qiang , Wang Shixi , Wu Mingyu 
Published: November 2022

Abstract
The 400MWt district-heating reactor (DHR400), which is one type of small modular reactor (SMR), is designed to solve district-heating problem in northern China, and its reactor characteristics and parameters are described in the paper. Reactor core inventory, accident source term and dose consequence of DHR400 is calculated in the event of fuel handling accident and core damage accident. The ORIGEN2 code developed by Oak Ridge National Laboratory is used to calculate the core inventory for an operating period of five years before refueling. Based on regulations for environmental radiation protection of China and regulatory guide of NRC, the accident scenario and source term assumption of design based accident and beyond design based accident are discussed. The accidental releases of radioactive material from reactor is calculated and evaluated based on the ARCON96 program sponsored by the NRC. In this paper, an overall systematic analysis method, which includes key aspects in the calculation, such as hazard assessment, source term estimation and timing, dose assessment and so on, is proposed to determine emergency planning zone (EPZ) of DHR400. The result shows that 114m could be used as the recommended boundary of EPZ of DHR400 in one specific DHR400 site.
Experimental Study on Statistical Properties of Two-Phase Flow Instability for Boiling
Authors: Sun Yuanli , Song Zhihao , Song Houde , Liu Xiaojing , He Hui 
Published: November 2022

Abstract
The study of two-phase flow instability is of interest in all fields encountering flow boiling. Compact systems relying on flow boiling due to its superior transport capabilities of heat and mass are prone to several modes of flow instability, which adversely affect the safety and performance of systems. As such, understanding of the flow instability is critical to the design and operation of compact flow boiling systems. In current paper, the instability caused by channel-to-channel interaction is investigated experimentally in a test section with 6 parallel small rectangular channels having hydraulic diameter of Dh = 526.2 μm by using deionized water as working fluid. Boiling leads to very asymmetrical flow distribution within the 6 channels, which results in the simultaneous existence of different flow regimes along the transverse direction of test section. Time traces of pressure drop for different channels and operation conditions are analyzed to explore the properties of channel-to-channel interaction by using spectral analysis. Bubble dynamics including nucleation and growth give rise to flow fluctuations in individual channel, and the time intervals between flow fluctuations are caused by a series of time-scale self-convolutions of bubble duration. Based on the aforementioned mechanism, the time intervals between flow fluctuations can be predicted statistically by the gamma distributions.
Application and Prospect of Additive Manufacturing Technology in Nuclear Fuel Assembly Fabrication
Authors: Ding Jie , Zheng Yixiong , Ye Chen , Li Cong 
Published: November 2022

Abstract
With the advent of additive manufacturing technology and the rapid development of technology in recent years, the design ideas and manufacturing methods of special-shaped and complex components have been greatly broadened. Combined with the continuous improvement of digital design and manufacturing tools, additive manufacturing technology has been widely used in aviation, aerospace, automotive and other industrial fields. Due to the high safety and reliability requirements for nuclear power equipments, the application of additive manufacturing technology in this field is still in the trial application stage before large-scale engineering promotion. This paper focuses on nuclear fuel, which is the core equipment of pressurized water reactor nuclear power plants. Based on the analysis of current domestic and foreign application examples, the key technologies of additive manufacturing of nuclear fuel are preliminarily explored, and the problems and challenges in the future engineering applications are prospected.
Research on Heat Transfer Performance of Heat Pipe in Mobile Small Reactor
Authors: Lu Huaichang , Zhou Tao , Liu Wenbin , Tang Jianyu , Gao Tianyu 
Published: November 2022

Abstract
Mobile heat pipe small reactors, with their simple structure, high mobility and inherent safety features, is a hot frontier topic for research at present. Based on Fluent software, numerical simulations are used to study the changes in the steady-state operating parameters of a high-temperature potassium heat pipe, and to investigate the effects of the horizontal acceleration, inclination angle and saturation temperature of the mass inside the heat pipe on the overall heat transfer performance of the heat pipe. It is found that the heat transfer performance of the heat pipe increases as the saturation temperature of the mass inside the tube decreases; the overall thermal resistance of the heat pipe tends to decrease and then increase slightly with increasing horizontal acceleration; the overall thermal resistance of the heat pipe tends to decrease and then increase with increasing tilt angle.
Fluid-Thermal-Mechanical Coupling Analysis of the Reactor Vessel of Natural Circulation Lead-Cooled Fast Reactor SNCLFR-100
Authors: Duan Wenshun , Zhang Kefan , Chen Hongli 
Published: November 2022

Abstract
Lead-based reactors mostly adopt integrated pool structures to simplify the design of reactor and have high inherent safety. As an important safety facility in the primary circuit, the reliability of the reactor vessel is critical for the safe operation of the whole reactor. The reactor vessel of natural circulation lead-cooled fast reactor SNCLFR-100 was taken as the object, and the corresponding simplified model was established. Based on ANSYS, the fluid-thermal-mechanical coupling analysis was carried out under the steady-state to study the influence of flow field and temperature field on the reactor vessel, and to reveal the distribution law of key thermal and mechanical parameters under the steady-state operation in this paper. The results of mechanical calculations were evaluated based on American society of Mechanical Engineers (ASME) standards, and the preliminary design scheme of reactor solid structure was optimized. The results showed that the stress evaluation results of the SNCLFR-100 reactor vessel meet the requirements of ASME standards in steady-state operation conditions. After optimizing the original solid structure, the stress evaluation results show that the maximum stress value was reduced by 43.3%, and the stress concentration was obviously suppressed, which indicates that the reactor container has a higher safety margin under the optimization scheme. The Steady-state calculation based on normal operation conditions provides a certain research basis for the subsequent transient calculation under accident conditions.
Design and Validation on Preload of Manway Bolt for Pressurizer of Nuclear Power Plant Under Hydrotest Condition
Authors: Chen Tao , Xiong Guangming , Fu Guanhua , Zhao Xiaolong , Wang Defu , Zhang Yongqiang , Liang Enming 
Published: November 2022

Abstract
Bolted-flange connected structure is widely used on Primary Equipment of Nuclear Power Plant and it is one of the most important assemblies of Primary Coolant Loop. Special attentions should be paid for the design and validation of this kind of structure for high risks of leakage and structure integrity of the equipment. In this paper, preload of manway bolt for Pressurizer of Nuclear Power Plant under hydrotest condition was calculated theoretically. Bolt-load distribution for the bolted-flange connected structure was simulated by Finite Element Analysis method with the above preload applied. Bolt load loss and asymmetric distribution during the test was presented resulting from the simulation for saddle-shaped manway flange. Assessment was carried out according to ASME in the aspect of stress in the bolt shank, pressure on the gasket and rotating angle of the flange (sealing verification) with the corrected preload applied. The simulation results were validated by 1:1 prototype test carrying out for the Pressurizer and they were consistent with the test results. The value of coefficient Ks for bolt preload design formula was confirmed by the test. Key aspects of influence on manway bolt preload design were summarized and full design formula as well as value of the design preload for hydrotest condition were given.
Neutronics comparative analysis of fuel assemblies using CSG and DAGMC methods
Authors: Sun NingYan , Liu ShiChang 
Published: November 2022

Abstract
The traditional Monte Carlo method uses the Constructed Solid Geometry (CSG) method to build up the geometry by creating surfaces and using Boolean operators. This method has obvious deficiencies for new type reactors with complex geometry and large size. This method has the problem of geometric overlap. Direct Acceleration Geometry Monte Carlo (DAGMC) method based on CAD geometry modeling can accurately describe higher-order surfaces, which can effectively overcomes the shortcomings of CSG modeling method and effectively build up the model. Additionally, CAD geometry has good visual effect. On the one hand, it is conducive to the verification of the correctness of the model. On the other hand, it is beneficial to the analysis of the calculation results DAG-RMC function based on Reactor Monte Carlo code RMC can directly use CAD model to carry out particle transport calculation. Based on CSG and DAGMC, this paper constructed three dimensional rod repeat structure fuel and plate type repeat structure fuel assemblies respectively and calculated keff as well as power distributions in order to verify the correctness of DAGMC in particle transport and critical calculation for complex geometry. Through comparisons, the keff and power distributions calculated by the two methods are in good agreement, verifying the correctness of DAG-RMC function.
The BDB-NS Hydrodynamic Model for Porous Spent Fuel Storage Racks Solved by Fourier Spectral Method
Authors: Liu Yu , Zhu Yuxuan , Zhao Fei , Liu Dong , Lu Daogang 
Published: November 2022

Abstract
Spent fuel storage racks (rack) are used to store highly radioactive spent fuel. Its seismic design is of great significance for radioactive containment. To reduce the transmission of seismic load, the fuel storage racks (racks) are in a free-standing design, whose bottoms are not fixed with the ground of the spent fuel pool wall. In an earthquake, the rack should not deform significantly to avoid the criticality of the fuel assembly during storage, and the water effect shall be considered. The porous rack design is widely used in the third generation PWR. Under seismic conditions, there is fluid penetration in the porous rack and flow around the outside. If the traditional fluid-structure coupling model of the closed structure is adopted, the analysis is too conservative, and the calculation method of transient CFD fluid-structure interaction parameters based on fine modeling is too time-consuming. To improve the calculation efficiency of fluid-structure interaction parameters of porous racks, a BDB-NS coupling model is established for the fluid immersed porous square cylinders. Combined with the Fourier Spectral Method and solved by MATLAB, the fluid force of the porous racks are obtained, and the parameters such as added mass and added damping are obtained. The algorithm is validated by a solid square cylinder in an infinite water zone the CFD results. This paper can provide a basis for the selection of fluid-structure interaction parameters of spent fuel storage racks for CAP1400 and HPR1000.
The Development and Application of Real-Time Risk Monitor System Based on Level 2 PSA for NPP
Authors: Li Wenjing , Liu Yu , Wei Wei , Yu Xinli 
Published: November 2022

Abstract
The method of real-time risk monitor system (RM) development based on Level 2 probability safety analysis (PSA) is established and a real-time risk model for a nuclear power plant (NPP) in China is developed in this paper. The risk monitor system can calculate not only the core damage frequency (CDF) but also the large early release frequency (LERF) with different plant configurations. In this paper, the application of risk monitor system will be discussed in two aspects: Configuration risk management and risk-informed technical specification (TS) optimization. Configuration risk management method based on risk monitor system is studied and case analysis is carried out for the NPP, which will improve the plant risk management ability and safety. Risk-informed technical specification optimization approach based on RM is studied and case analysis is performed for the TS, which can prolong the completion time for a single equipment/system failure in TS to improve the operation economy of NPP. In addition, the evaluation of completion time for multiple equipment failures is the supplement to the traditional TS. The Level 2 PSA risk monitor system development is a good engineering practice of Level 2 PSA in NPP, and is of great significance for improving both the safety and economy of the plant.
Numerical Simulation Analysis of Silicon Carbide Whiskers Doped Uranium Dioxide Composite Fuel for High Thermal Conductivity
Authors: Lu Linyuan , Guo LiHua , Zhong Yajuan , Zhang Feng , Lin Jun 
Published: November 2022

Abstract
SiCw is expected to be candidate material for Accident Tolerant Fuel (ATF). Nevertheless, the distribution, morphology and size of SiCw have impacts on the thermal conductivity of composite fuels. An analysis model for UO2-SiCw was established using finite element method (FEM), and the effects of whisker orientation, size, volume fraction and aspect ratio on thermal conductivity were analyzed. The generation algorithm of SiCw was improved, and the results of thermal conductivity calculations showed that the improved model is closer to the experimental value. The effect of whisker orientation on thermal conductivity can be calculated by the cosine law of zenith angle and azimuth angle, and deriving a model for calculating thermal conductivity with specific whisker orientation. With the decrease of whisker size, the influence of Kapitza thermal resistance is significant. The formula used to calculate the whisker threshold diameter was deduced, and the results are in good agreement with the FEM results. With the increase of volume fraction of whiskers, the thermal conductivity increases linearly. For morphology, whiskers with high aspect ratio can achieve higher thermal conductivity. There is a value for the ratio of whisker length to matrix size, above which increasing whisker length will slow the increase in thermal conductivity (0.3 for whiskers along the direction of thermal conduction). The thermal conductivity of UO2-SiCw was optimized using the above conclusions, and the radial temperature distribution of pellet was analyzed. The results show that taking full advantage of the anisotropy of whisker heat transfer can significantly improve the thermal conductivity of composites and improve the operation condition of the pellet. The results can provide a good theoretical basis for designing whisker-doped composites with high thermal conductivity.
Hydrogen Release Reaction From Sodium Hydride With Different Sample Quantities
Authors: Doi Daisuke 
Published: November 2022

Abstract
In sodium-cooled fast reactors (SFRs), hydrogen is a major nonmetallic impurity in the coolant during normal operation. A higher hydrogen concentration than the gas-liquid equilibrium had been transiently detected in the gas space of the actual SFR plant. However, the chemical reactions that caused hydrogen generation, which involve several sodium compounds, have not been identified. Furthermore, the thermal behavior of these hydrogen release reactions has not been thoroughly investigated. In this study, the hydrogen release behavior of sodium hydride, which could be involved in all of these reactions, was clarified by two experimental methods dealing with different sample quantities. In the thermal analysis with a semi-micro sample of about 1mmol, the hydrogen generation was demonstrated by mass spectrometry as the sample mass decreased, suggesting thermal decomposition. A monomodal hydrogen release curve similar to the thermal analysis result was obtained in the heating experiment with a macro amount sample of about 1mol. These experimental results showed consistent activation energies within the standard error. Therefore, it was elucidated that the ideal reaction behavior obtained by thermal analysis could be sufficiently extrapolated to the reaction behavior occurring in a larger amount of sample. These findings provide fundamental insights into the thermal decomposition of sodium hydride and are indispensable for analyzing hydrogen release behavior in other hydrogen release reactions involving sodium hydride.
Determination of Radiation Shielding Properties of Low-Density Polyethylene Reinforced With Bismuth Against Gamma-Rays
Authors: Abd Rahman Nur Shafriena Farhana , Sazali Muhammad Arif , Ripin Azuhar , Mohd Ali Nur Syazwani , Sarkawi Muhammad Syahir , Jamaluddin Khairulnadzmi 
Published: November 2022

Abstract
Radiation hazard has inevitably become a global threat due to the widespread use of radiation technology in industrial manufacturing and medicine. As in medicine, the workers are in danger of being exposed to high-penetrating gamma-rays during the treatment of cancerous tumours. Hence, developing such an excellent shielding material from the polymer is vital since it can be used broadly, is low cost, and is non-poisonous to nature. This study aimed to observe the effect of additive content such as bismuth on shielding performance. To test the hypothesis that high-density material leads to higher radiation shielding performance, fabricated bismuth-low-density polyethylene (LDPE) composites that have been processed via internal melt-mixing and compression-moulding processes were labelled with LDPE-100, Bi-5, and Bi-10 containing 0, 5, and 10 wt.% of bismuth content with a thickness of 0.1, 0.2, 0.3, and 0.4 cm respectively. Throughout the experiment, a photon from a gamma source of 241Am and 133Ba with an energy range of 59 keV and 356 keV were captured by a solid-state detector to calculate the count rate in count per second (cps). The fundamental shielding parameter such as mass attenuation coefficient, half-value thickness, tenth-value thickness, and mean free path was also computed using Phy-X software to evaluate the radiation shielding effectiveness. The results showed that Bi-10 polymer composites with 10% wt. of bismuth filler had MAC, HVL, TVL, and MFP values of 1.159 cm2/g, 0.554 cm, 1.838 cm, and 0.798 cm, respectively, and may be used as an effective shielding material for gamma radiation with an energy of 59 keV compared to 356 keV.
Research on Operation Life Extension Time Limited Aging Analysis Key Method for Electrical Penetration Assembly of Nuclear Power Plants
Authors: Wang Binfeng , Chen Zhilin , Huang Ping , Shi Zhenhao , Zhang Yanwei , Liu Tao 
Published: November 2022

Abstract
Environmental qualification (EQ) is a process to ensure that equipment will be capable of withstanding the ambient conditions and executing specific function that the equipment is actually called upon to be performed under accident conditions. The main purpose of EQ is to provide reasonable evidence to ensure that common cause failures will not occur. Electric penetration assemblies (EPAs) belong to EQ components and perform safety-related functions. The time limited aging analysis (TLAA) of EQ components includes all long-lived, passive and active electrical components and commodities located in a harsh environment that are important to safety. This paper takes the electric penetration assembly (EPA) of a early domestic M310 NPP as the research object, based on the components, materials, environment, aging mechanisms and effects of EPA, combined with domestic and foreign standards, specifications and latest research results of equipment qualification and operation license extension, the time limited aging analysis key method of EPA of a early domestic M310 NPP is studied, the test scheme is designed, the test sample is selected, the key parameters are determined, the supplementary environmental qualification of EPA of a early domestic M310 NPP is finished. At the same time, combined with the current management status of EQ equipment at home and abroad, some management suggestions are proposed to maintain the validity of the qualification status during the period of operation license extension.
Mechanical Vapor Re-Compression Evaporation Treatment of Radioactive Waste Water
Authors: Qiaofen Liu , Ming Huo , Fei Gao , Linjun Yang , Yong Liu 
Published: November 2022

Abstract
Evaporation method has been widely used in radioactivity wastewater treatment of nuclear power plants because of its high decontamination factor (DF) and wide application range of water quality. However, the traditional evaporation system has high energy consumption. Aiming at the recovery of latent heat of secondary steam during stable evaporation of wastewater in radioactivity, a new evaporation system & mechanical vapor recompression (MVR) was proposed and studied. Firstly, a mathematical model is established according to the conservation equations of mass and energy. The factors affecting energy consumption, decontamination factors, and secondary waste that needing attention due to radioactivity are discussed. According to the improved scheme, a prototype was built and the wastewater treatment experiment in simulated radioactive water was carried out. The actual energy saving rate and decontamination factor are calculated and analyzed. The results show that, at the conditions of evaporation rate of 600kg/h and evaporation temperature of 100°C, compared with traditional evaporation, MVR can save 84.8%energy, and the decontamination factor (DF) reaches 105 in simulated waste liquid treatment experiment. MVR shows that this method has obvious advantages in economic and environmental benefits.