bluuebunny/crossref_metadata_2025_split huggingface dataset:
A Detailed Model to Predict Mechanical Characteristics of Fuel Assembly
Authors: Zhao Jingjing , Xu Shiyin , Pan Wenan , Sun Jingya , Du Yuxi 
Published: November 2022

Abstract
As the core component and the first safety barrier of a nuclear power plant, the structural integrity of fuel assemblies (FAs) is an important factor in ensuring the safety of nuclear power. Safety analysis and evaluation of fuel assemblies under seismic conditions is a requirement of nuclear power plant design codes and safety reviews. The mechanical properties of a single FA determine the operational and seismic performance of the core fuel assembly. Due to the complexity of the FAs in terms of structure and boundary conditions, test is the most common method to obtain the mechanical properties. In order to predict the mechanical properties of the fuel assembly, a nonlinear finite element detailed model was developed. The contact and frictional behavior between the fuel rod cladding and the spring/dimple convexity were considered in the modeling to simulate the nonlinear behavior of the fuel rod better. To verify the validity of the model, virtual tests of lateral stiffness test, forced vibration test and impact test were conducted on the model. The comparison with the experimental results shows that the model can reflect the mechanical properties of the fuel assembly well. The modeling method and calculation conclusions in this paper are of guidance for fuel assembly design.
Access Time and Safety Benefit Analysis of Additional Temporary Water Injection Measures on Primary After the Fukushima
Authors: Liu Weidong , Wang Zhao , Liu LeiLei , Zhang Kuan 
Published: November 2022

Abstract
After the Fukushima nuclear accident, National Nuclear Safety Administration carried out plant inspection nuclear safety on the nuclear plant in operation and under construction, and put forward improvement requirements for each nuclear power plant according to the inspection results. One of the improvement actions was to add temporary water injection measures on the primary side. At present, all power plants have plant requested to add mobile pumps and related water supply pipelines and operating procedures. By summarizing the calculation cases of plant, which selected representative accident sequences for deterministic analysis in the past, this paper determines the effect of implementing temporary water injection measures in primary circuit on accident mitigation at different times, and the results can provide time suggestions for the preparation of mobile equipment in plant. At the same time, the safety benefits after implementing temporary water injection measures in primary circuit in plant are analyzed by using probability analysis method. Based on the analysis of deterministic and probability theory, and considering the preparation time of accessing mobile equipments, it is considered that 4 hours is a reasonable access time for mobile equipments under SBO accident conditions. In the actual access process, this time can ensure that operators have a more suitable time to complete the preparation work of mobile equipments in operation, and at the same time contribute to the reduction of large release frequency. At the same time, it should be noted that at present, the contribution of implementing temporary water injection measures on the primary side under SBO accident is mainly analyzed. In fact, some accident situations with slow development except SBO accident also have great access possibility, so the actual safety benefits of temporary water injection measures on the primary circuit may be more.
Design and Optimization of a PGNAA System Based on D-D Neutron Generator for Cement Composition Analysis
Authors: Chen Si-Yuan , Li Guang-Hao , Jia Shao-Lei , Lu Zhao-Hu , He De-Dong , Ke Gong , Shi Guang-Yu , Jing Shi-Wei 
Published: November 2022

Abstract
Online composition analysis is a crucial method for energy-saving and environment-friendly cement industry. A prompt gamma neutron activation analysis (PGNAA) system, which is aimed to determine the elemental concentrations of cement sample, has been designed and optimized by the Monte Carlo approach. This system consists of a 2 × 2” Bismuth Germanate (BGO) scintillation detector, a portable D-D neutron generator, a multi-channel analyzer, the reflector, the moderator and the shield. Out of consideration for minimization of costs and maximization of the thermal neutron flux at the output surface, different models of reflectors and moderators were tested in the MCNP code. Meanwhile, lead shield was also designed to reduce gamma-ray background produced by the nature and the system itself. In addition, simulated spectra of a standard cement sample were obtained to test the performance of this system. For ensuring the safety of the system operators, the dose rate distribution around the system were evaluated. The results demonstrated the feasibility of adopting this optimized design for chemical composition analysis of cement sample, and provided data reference for similar research. Eventually, an in-situ system has been established for further study.
Multiphysics Analysis of Fuel Performance and Tritium Migration in FeCrAl and Cr-Coated Zircalloy Cladding Under PWR Normal Operating and Transient Conditions
Authors: Liu Rong , Yang Liwen , Liu Shengyu 
Published: November 2022

Abstract
Due to the fact that the Cr-coated Zircalloy and the FeCrAl alloy have excellent high-temperature oxidation resistance, corrosion resistance, they are considered to be potential replacement materials for PWR cladding. However, it is also reported that the poor tritium resistance of FeCrAl alloy will accelerate the penetration of tritium throughout the cladding and increase the treatment cost. In this work, the physical models and tritium migration models of Cr-coated Zircalloy cladding and FeCrAl claddingare reviewed firstly. Then, based on the developed fuel performance analysis code CAMPUS, the fuel performance models of Cr-coated Zircalloy cladding and FeCrAl cladding are implemented, and a new model of the tritium migration in different claddings is implemented too. Finally, the performances of three fuel cladding combinations (UO2 pellets and Zircalloy cladding, Cr-coated Zircalloy cladding and FeCrAl cladding) under normal and LOCA conditions are simulated and discussed. And the tritium resistance performance of different fuel cladding combinations is further calculated and analyzed with the new tritium migration model. The calculation results show that under normal and LOCA conditions, compared with Zircalloy cladding, the application of FeCrAl cladding can reduce the overall temperature of fuel and improve fuel safety margin. And the application of Cr-coated Zircalloy cladding and FeCrAl cladding can effectively delay the failure time of cladding under LOCA condition; In terms of tritium penetration, the Cr-coated Zircalloy is found to have better tritium resistance effect than FeCrAl alloy, which greatly reduces the pollution treatment cost of reactor operation.
Quantitative Probabilistic Safety Analysis of Locking Mechanism of Nuclear Power Control System Products
Authors: Jiang Zhirui 
Published: November 2022

Abstract
Nuclear power control system products must be firmly fixed in the cabinet to ensure that the control system can work normally under earthquake and other working conditions. The locking mechanism of a nuclear power control products were analyzed by FTA, and the quantitative probabilistic safety analysis of the key failure mode locking mechanism loosening event was carried out. Firstly, the Finite element model of locking mechanism was established by using ANSYS, conduct a deterministic simulation analysis under seismic conditions. Then, based on the reliability theory, the failure mode interference equation was established according to the deformation of the locking hook of the locking mechanism, Monte Carlo reliability algorithm program and ANSYS joint simulation interface program were developed, for the failure of locking mechanism loosening, Monte Carlo sampling statistical joint simulation calculation was carried out by using the ANSYS joint simulation interface program, the probability of loosening failure event of locking mechanism under different confidence was calculated. This paper explores a new method for quantitative probabilistic safety analysis of locking mechanism, which has reference value for quantitative probabilistic safety analysis of similar mechanisms. The locking mechanism loosening event probability in the fault tree of probabilistic safety analysis was accurately calculated by using this method, through the introduction of reliability algorithm, it focuses on the evaluation core processing of uncertainty, which is an improvement of probabilistic safety analysis method.
Flow-Solid Coupling Analysis of Pressure Vessels on Floating Nuclear Power Plants
Authors: Wang Yuchao , Lu Daogang , Zhao Fei , Liu Yu 
Published: November 2022

Abstract
Lead-cooled fast reactors are an important choice for small-scaled reactor in floating nuclear power plants. Compared with land-based nuclear power plants, floating nuclear power plants have advantages in stability, which eliminate the impact from earthquake. One aspect remains to be solved is that the lead-cooled fast reactors contains a large volume of liquid that may slosh under the marine environment. The sloshing of liquid may cause external pressure on the nuclear reactor vessel and internal components., which has a great impact on the safety of the reactor. To study the fluid structure interaction between liquid and reactor vessel, this paper establishes an numerical model for the mechanical analysis of the main vessel of a floating nuclear power plant under loads of the marine environment. The mechanical response of the main vessel of a floating nuclear power plant in the marine environment is analyzed by applying the dynamic mesh method and VOF model. The motion of marine environmental loads are applied by the remote displacement method. We study several typical loadings in operation conditions. The study shows that the stress on reactor vessel caused by fluid sloshing is relatively small when long-period loading is applied; and the effect of swaying is larger than the rolling.
The Study of Coupled Fuel Performance Analysis and Neutronics With COMSOL and RMC
Authors: Liu Zhenhai , Qi Feipeng , Zhou Yi , Li Quan , Chen Ping , Chen Hao , Ma Chao , Zhao Bo , Li Chenxi 
Published: November 2022

Abstract
Predicting fuel behavior in the reactor is a typical multi-physics coupling problem. The traditional fuel performance analysis adopts a decoupling way. To provide a higher fidelity tool for fuel performance simulation, the multi-physics coupling software COMSOL and 3D Monte Carlo neutron transport code RMC are coupled. The fuel rod thermo-mechanical coupling analysis module is constructed in COMSOL. RMC is innovatively wrapped as a component of COMSOL to communicate with the fuel performance module. Grid mapping algorithm is set up using COMSOL’s functionality. A carefully designed segregation solver sequence is used to control the coupling process. The fine power distribution calculated by RMC is transferred to COMSOL fuel thermal calculation, and the fuel temperature and the coolant temperature and density calculated by COMSOL are transferred to RMC. In a Picard iterative way, the fuel rod lattice level neutron thermomechanical tight coupling is realized. The coupling is demonstrated in a fuel rod BOL steady state case. The fuel performance analysis is enhanced by the higher fidelity neutron flux solution.
Study on Steady Operational Characteristics of Floating Nuclear Reactor Under Ocean Condition
Authors: Cheng Kun , Chu Xiao , Qiu Zhifang , Xian Lin , Deng Jian 
Published: November 2022

Abstract
Different from land-based reactors, floating reactors are affected by the ocean condition. Numerical simulation is conducted to study the effects of heaving and rolling motion on typical floating nuclear reactor (FNR) under rated power operation by ARSAC code with ocean condition calculation module. The calculation results have shown that heaving motion can cause the periodic fluctuations in the flow rate of both the primary and secondary loops in FNR. The amplitude of heaving motion is the main influence factor on the flow rate fluctuation of the primary loop while the effect of heaving motion period is limited. Rolling motion can cause the out-phase flow fluctuation in the two symmetric loops of the primary system. However, the rolling motion has limited impact on the core flow rate and cannot cause dramatic changes of system parameters at the core. Within the range of motion parameters used in this paper, heaving and rolling motion cannot have a substantial impact on the safe operation of FNR under rated power operation.
Effects of Aggregation on Gravity Settlement of Submicron Aerosols Under High Temperature and High Humidity Conditions
Authors: Chen Jun Yan , Gao Pu Zhen , Gu Hai Feng , Yu Hui Yu 
Published: November 2022

Abstract
Coupling aerosol gravity sedimentation mechanism and aggregation effect, considering the correction of submicron particle free path and gas dynamic viscosity in thermal environment, a gravitational sedimentation model suitable for submicron aerosol accompanied by coalescence effect under high temperature and high humidity conditions is established, and the established model is verified by the results of thermal experiments. The aerosol aggregation model is established using the discretization method. By controlling whether the aggregation mechanism occurs in the established model, the influence of aggregation on gravitational settling when steam fraction and ambient pressure are changed is studied. In the study, it was found that the aggregation effect has a nearly 1-fold acceleration effect on the decay constant of the number concentration of submicron aerosols during gravitational sedimentation, and the acceleration effect on the decay constant of the mass concentration can be ignored, but there is an acceleration trend. When submicron aerosols settle under high temperature and high humidity conditions, the number median particle size tends to increase, that is, the increase of the number median particle size by the aggregation effect is greater than the decrease of the number median particle size by gravity sedimentation. After considering the aggregation effect, the number concentration decay constant and the mass concentration decay constant when the submicron aerosol undergoes gravitational sedimentation will decrease with the increase of the ambient pressure, and increase with the increase of the steam fraction. However, the number concentration decay constant changes larger, the mass concentration decay constant has a trend of change, and the change range is relatively weak. In addition, in the sensitivity analysis of the aerosol coalescence and gravitational sedimentation models, it is found that the molecular free path has a greater impact on the calculation results of aerosol coalescence and deposition in a thermal environment than the dynamic viscosity. Moreover, the influence of Brownian coalescence on the aerosol particle size spectrum is much greater than that caused by gravitational coalescence.
Development and Application of Modular Information Management System for Nuclear Power Engineering
Authors: Gong Xiaocheng , Xu Xinwei , Zhou Wei 
Published: November 2022

Abstract
Modular construction technology is one of the effective ways to shorten the construction period and reduce the cost of nuclear power plants. It is one of the important features of the development of nuclear power engineering construction technology. The application of the module in China has a history of more than 20 years, and it was first used in the construction technology of nuclear island workshop dome steel lining. As modularization is the technology of integrated design,procurement and construction,and is based on the three-dimensional design platform, the close cooperation among design, procurement and construction also needs the support of the corresponding information platform. It is more urgent and necessary for the unified, shared and rapid interface response processing of upstream, downstream and management information. Based on the actual needs of nuclear power project construction and relying on the modular three-dimensional model, this paper develops and constructs a three-dimensional visual modular comprehensive information management system integrating the business data of design, procurement, construction, commissioning and management, realizes the efficient coordination and fine management of modular advanced construction technology of nuclear power project, and provides an information management platform for modular construction of nuclear power project.
Study on the Added Mass of the Double-Fixed Constrained Flexible Coaxial Cylindrical Shells With Liquid in the Narrow Gap
Authors: Li Donghao , Lu Daogang , Zhu Yuxuan , Liu Yu 
Published: November 2022

Abstract
There are many coaxial cylindrical thin-wall structures with a narrow fluid gap in the reactor block of the pool type sodium cold fast reactor (such as primary pump support and its thermal shield). The fluid-structure interaction (usually represented by added mass) of shells and the fluid gap changes the structure’s dynamic characteristics under seismic conditions. Therefore, it is significant to calculate the added mass of the coaxial cylindrical shells. At present, there are few studies on the vibration modes and added mass of flexible shells with double-fixed constrain. For the flexible coaxial cylindrical shells model, due to the coupling of the inner and outer shell, the vibration modes of the inner and outer shell in the same and opposite directions are generated. In this paper, the added mass of two vibration modes in the same direction and in the opposite direction are investigated. The finite element models (with water and without water) of the coaxial cylindrical shells were established using ANSYS software, and the vibration modes and added mass were calculated by FEM. Further, the effect of different outer shell rigidity on the added mass of coaxial cylindrical shells was studied by changing the material parameters of the outer shell.
Fault Diagnosis Based on KPCA and RUSboost for Nuclear Power Plants
Authors: Wang Yiqing , Sun Peiwei 
Published: November 2022

Abstract
Nonlinear relationships among different process variables are very common in nuclear reactors. When constructing a nonlinear model to for process monitoring, linear-based monitoring technique has been proved inefficient and problematic. Kernel principle component analysis (KPCA) combines the principle component analysis and kernel method to cope with fault diagnose in nonlinear system. Experiments are conducted using data samples from Generic Pressurized Water Reactor Simulator(GPWR) to simulate sensor fault and three types system faults. The nonlinear relationship among monitoring variables is analyzed. KPCA is used to capture the nonlinear relationships to distinguish the sensor fault and system fault, which shows the superiority of KPCA to detect a univariate fault. Subsequently, system faults are further analyzed as a classification problem.
Samples imbalance is another common problem existing in diagnose domains. When training samples of one class outnumber than other class, traditional classification algorithm performance may deteriorate. Random under-sampling boosting (RUSboost), a hybrid sampling and boosting algorithm is used to alleviate the imbalance problem. The performances of classifiers by four metrics is evaluated. When compared to Adaboost (contains only boosting), tree variants (base model and ensemble bagged trees) and Gaussian SVM, RUS boosting shows significantly high accuracy and robust performance in dealing with different ratio of training samples, which shows that RUSboost is an excellent technique to cope with imbalanced data in fault diagnose for nuclear power plants.
Investigation on Irradiation Resistance of Anti-Galling Lubricant for High Temperature Fasteners of Nuclear Power Plant
Authors: Liu Xiao Qiang , Jiao Ming , Shi Xiu Qiang , Bao Yichen 
Published: November 2022

Abstract
Lubricant grease is a special material commonly used in the field of nuclear power plant. It is widely used in the sliding fit and press fit parts of bolts, studs and valves to prevent thread wear, seizing as well as stress corrosion cracking of fasteners during assembly, service and disassembly. In this paper, the performance of lubricant materials such as wear resistance and high temperature oxidation resistance under irradiation environment were studied. The results show that the lubricant grease does not change significantly and there is no grease stratification after irradiation with 0.1MGy integrated irradiation dose. However, after irradiation with 10MGy integrated irradiation dose, the irradiation has an obvious destructive effect on the oxidation resistance performance of the grease, and the grease appears obvious stratification. With the increase of irradiation dose, the thermal decomposition temperature first decreases and then increases. Under 0.1MGy irradiation dose, the thermal decomposition loss of the sample changes little, but at 10MGy irradiation dose, a large amount of mass loss occurred, up to more than 80%. The four-ball friction experiment shows that its extreme pressure resistance improves with the increase of irradiation dose, and the sintering load increases. These data provide technical guidance for the long-term reliability of nuclear power fastener materials in service.
Effect of Hydration Time on the Hydro-Mechanical Behavior of Compacted GMZ Bentonite With an Artificial Annular Gap
Authors: Luo Hewen , Ye Weimin , Wang Qiong , Xu Libo 
Published: November 2022

Abstract
According to the conceptual design of high-level radioactive waste (HLW) disposal repositories, the buffer barrier will be constructed by compacted bentonite blocks, which will inevitably produce technological gaps (voids) in the repository. During the consequent long-term operation of the repository, the technological gaps will be gradually filled up on hydration of the bentonite blocks with infiltration of underground water from the host rock. This hydration process will be accompanied by variations of the hydro-mechanical behavior of the buffer system. Therefore, it is necessary to study the changing process of the hydro-mechanical behavior of bentonite blocks with consideration of influences of the technological gaps. In the present work, a series of hydration tests were carried out on compacted GMZ bentonite specimens with initial annular gaps. During the hydration, variations of swelling pressure and hydraulic conductivity were measured. After each test, distributions of dry density and water content in the specimen were determined and the microstructure at different positions was investigated. Results show that, the swelling pressure of the specimen increased rapidly first and then decreased slightly during the hydration process, while the hydraulic conductivity kept decreasing with time. Moreover, closure of the technological gaps induced heterogeneity of the bentonite blocks, which was characterized by the decrease of dry density and the increase of water content with increasing distance to the center of the specimen. With the hydration time elapsed, the bentonite blocks could be divided into compression zone and swelling zone along the radius. In the compression zone, the dry density of the specimen gradually increased, with a decrease in the total pore void ratio and the macro-pores void ratio. In the swelling zone, the dry density decreased with an increase of the total pore void ratio and the undetectable pore void ratio.