bluuebunny/crossref_metadata_2025_split huggingface dataset:
Manufacturing Process of Fuel Salt Discharge Tank for 2Mwt Liquid Thorium Based Molten Salt Reactor
Authors: Wang Yan , Shi Hong 
Published: November 2022

Abstract
At present, the research on Thorium Based Molten Salt Reactor (TMSR) mainly focuses on molten salt-molten salt heat exchanger, molten salt-air heat exchanger, high-temperature molten salt pump, gas-liquid separator and Hastelloy structure and performance, and there is no research on the manufacturing process of fuel salt discharge tank. The fuel salt discharge tank mainly performs the function of fuel loading and unloading, that is, loading and unloading the fuel salt through the gas pressure. The difference gas pressure between the core and the fuel salt discharge tank make the fuel flow into the reactor pressure vessel or flow out of the reactor pressure vessel. The fuel salt discharge tank also has the function of storing the molten salt fuel and preventing fuel leakage when the core is overheated. It exists as a safety facility and has a very important impact on the normal and safety operation of the reactor system. Based on the requirements of standards and specifications, the main manufacturing process of fuel salt discharge tank were studied, technological processes such as tank material selection, head forming, barrel rolling, welding and pneumatic test were introduced. The fuel salt discharge tank of liquid TMSR meeting the requirements of standards and specifications was successfully developed.
Research and Application of 3d Visualization Digital Twin System for Ultrasonic Testing of Key Equipment in NPP
Authors: Yang Zheng-Hui , Zhang Yi , Wang Hao , Zhang Ya-Nan 
Published: November 2022

Abstract
The internal conditions and surface defects of key equipment in nuclear power plant (NPP) are mainly detected though non-destructive testing (NDT) technology, which is the key basis for maintenance and diagnosis of nuclear power main equipment. At present, there are many types and huge quantities of NDT data. Due to the lack of unified information management means, it depends more on personal skills and experience. On the other hand, discrete data files and 2D detection data can’t provide the staff with intuitive and stereoscopic visual effects, which affects the prediction and analysis of the staff to a certain extent. In view of the above problems, this paper deeply analyzes the characteristics of nuclear power NDT data, and proposes a 3D visualization method for ultrasonic defects of NPP reactor pressure vessel (RPV) based on digital twin. It also forms a 3D visualization digital twin system, which overlays and maps the 3D digital model of RPV with the reconstructed 3D defects, while providing the necessary 3D interaction. Through the research results of this paper, the digital and intelligent management methods of NDT data in NPP have been improved, and an outstanding 3D basis for prediction and analysis of NDT defects in NPP has been provided.
Verification and Validation of the Fuel Assembly Bypass Flow Analysis Code PEONY V1.0
Authors: Chen Changyi , Yan Jingwen , Xi Yanyan , Zhang Yuxiang , Li Weicai , Wu Change , Lv Qifeng , Ran Shuangye 
Published: November 2022

Abstract
Fuel assembly bypass flow is the coolant flow through the guide thimble and the instrument. The fuel assembly bypass flow should be limited to ensure that there is sufficient coolant in the core channel to remove the heat released by fuel fission, and the bypass flow should be large enough to avoid the boiling of the coolant in the guide thimble and the instrument tube. The calculation of the fuel assembly bypass flow is also important for the mechanical analysis of SCCA (Stationary Core Component Assembly) hold down system and for the analysis of control rod drop. The PEONY V1.0 code developed based on pressure drop balance between different flow channel was used for calculating the flow distribution and then the hydraulic force for SCCA mechanical design verification. The PEONY V1.0 code is verified by STEP fuel assembly bypass flow test. In order to obtain a more accurate measurement results, pressure drop test result is used for flow rate correction. The verification and validation of PEONY V1.0 shows that the code can provide an reasonably conservative calculation for engineering analysis.
Propulsion Control of Marine Nuclear Reactor by Hardware-in-the-Loop Simulation (HILS)
Authors: Zhang Ao , Wang Xiang 
Published: November 2022

Abstract
Compared with offline simulation based on software completely, Hardware-in-the-loop Simulation (HILS) is a simulation method based on software and hardware coupling, which can not only verify the principle of the model studied, but also feedback hardware response for comparison, shorten product test cycle, improve research and development efficiency and accuracy. Taking the marine nuclear reactor as the background, this paper modelled the main circuit of energy power and the part of propulsion power output, simulated the energy power module offline based on Modelica language, and coupled Arduino MCU with LEGO parts to simulate the hardware response of propulsion power output. This study used OpenModelica to construct the steady-state and transient simulation operation of the off-line simulation model, realized the interaction between the hardware signal input and the offline simulation model, processed the mechanical response of the simulation output and communicated with the external hardware. Finally, the feedback action response was compared with the key parameters in the existing literature. The results show that the simulation results of the model and the hardware response feedback results are consistent with the literature parameters, and based on the coupling of Modelica language with Arduino MCU can carry out the HILS for the marine nuclear power system in steady state and transient state.
Design and Comparison of Drainage Technology Scheme for Multi-Purpose Modular Small Reactor Science and Technology Demonstration Project
Authors: Yujia Mao , Liyuan Sheng , Ruobing Yang , Wei Bai , Jing Li 
Published: November 2022

Abstract
Compared with other nuclear power projects, the layout and site conditions of the Multipurpose Modular Small Reactor Technology Demonstration Project (hereinafter referred to as “Small Reactor”) have obvious particularities, which makes the design of the drainage process of the small reactor demonstration project has special considerations compared with that of the dual-reactor nuclear power plant. It mainly includes the following aspects: first, because of the single-stack arrangement, the plant water system cannot use the facilities of other units to drain water during shutdown for maintenance; second, the impact on the already operating Units 1 and 2 needs to be considered; third, consider the interaction with the Units 3 and 4; fourth, the unit cost of small reactors is relatively high, and the construction cost of drainage facilities needs to be controlled. In this paper, two drainage technology schemes are formed according to the particularity of the small reactor and the geological conditions of the site, considering the feasibility of the structure, the results of warm drainage, the construction method and the cost are analyzed respectively, the countermeasures are studied for possible problems and a reasonable arrangement method is finally recommended, which realizing the feasibility, reliability and economy of drainage engineering of Small Reactor.
Study on Performance of New Flexible Shielding Materials and First Demonstration Application in CPR Nuclear Power Plant
Authors: Li Yulong , Liu Feng , Liu Xiajie , Li Li , Kang Zheng , He Ye 
Published: November 2022

Abstract
Excessive gamma rays will be emitted when a nuclear power plant is under the overhauling, leading to a certain number of hotspots. To meet the shielding requirements of these hotspots, a new type of flexible shielding materials with a “sandwich structure” has been developed. New flexible shielding materials consist of two parts, one is silicone rubber with excellent physical and chemical properties by adding reasonable multi-component additives as matrix, and the other is lead powder as shielding body. This kind of material with lead powder dispersed uniformly in silicone rubber was prepared through mixing process. Based on this, the article has carried on the discussion to the performance of functional layer of the new flexible shielding materials. The problem that the process stability of ultrafine powder distributed uniformly in matrix with various densities among different regions has been solved as well. Furthermore, this kind of material has been adopted in CPR nuclear power plant as the first demonstration project. The site shielding effect has been recognized by WANO (world association of nuclear operators). As a result of this research, it is possible that the scheme of radiation shield for a short or long time can be make quickly when nuclear power plant running or overhauling. Meanwhile, this study is of great significance for the popularization and application of shielding materials.
Three-Factor Winding Wire Insulation Aging Evaluation Based on Design of Experiment
Authors: Zhou Daofan , Yan He , Diao Xingzhong 
Published: November 2022

Abstract
Winding wire, as an important part of electrical equipment such as motor coils, is vital to the stable operation of reactor electrical equipment. The insulation aging of winding wire in reactor environment is affected by multiple factors such as temperature, voltage and irradiation. Due to the complex aging mechanism under the effect of multi-factor coupling, it has become a research hotspot to explore the influence of single factor and multi-factor coupling on the insulation lifespan of winding wire, and give the insulation lifespan of winding wire under multiple factors. In this paper, design of experiment (DOE) method was used to test design and results analysis, and the corona resistance tests of winding wire were carried out to get the corona resistance lifespans, which were regarded as the insulation indicator of winding wire under multiples factors. Full factorial experimental design was planned by changing the temperature, irradiation dose, and voltage amplitude at two level. The corona resistance lifespans of the samples were obtained, and the data were processed by the two-parameter Weibull distribution. Finally, the effects of single-factor and multi-factor coupling on insulation lifespan were discussed, and a model for evaluating the insulation lifespan of winding wire under the combined effects of three parameters was developed.
Preliminary Prediction of Young’s Modulus and Thermal Conductivity of Porous U-50 wt.% Zr Alloy: A Molecular Dynamic and Semi-Empirical Study
Authors: Cai Mengke , Cong Tenglong , Gu Hanyang 
Published: November 2022

Abstract
In the upgrades and innovations of nuclear fuel material, the U-50 wt.% Zr alloy is regarded as one of the most promising metallic fuel materials, due to its excellent thermal response, acceptable irradiation performance, and ease of fuel recycling. Under in-pile irradiation, large temperature gradients and dimensional changes contribute to complicated fuel thermal-mechanical behaviors, including the pore effect induced by fission gas production. However, the deficiency of the physical parameters of the porous U-50 wt.% Zr alloy makes it hardly possible to conduct fuel performance prediction under high irradiation conditions. To obtain Young’s modulus and thermal conductivity of porous U-50 wt.% Zr alloy, the molecular dynamics (MD) code LAMMPS and the modified embedded atom method (MEAM) potential for binary U-Zr system were incorporated. In this study, three-dimensional elastic constants were calculated by engineering strain loading method at different ambient temperatures and porosities, and the effective Young’s modulus was computed via Voigt averaging scheme. The phonon thermal conductivity was simulated with the Non-Equilibrium Molecular Dynamics (NEMD) method, and the electron thermal conductivity was predicted by semi-empirical correlations and existing density functional theory (DFT) results. The parallel model, series model and effective medium theory (EMT) were adopted to consider the mixture and pores effect. Finally, porosity factors were proposed to establish new semi-empirical correlations, which could give a preliminary prediction of Young’s modulus and thermal conductivity for porous U-50 wt.% Zr alloy.
Assessment and Control Value Formulation on Radiological Consequence of On-Site Worker due to Radioactive Component Failure for Pressurized Water Reactor Plant
Authors: Jing Zhou , Wen Wang Ran , Quan Gong , Wei Feng Lv 
Published: November 2022

Abstract
All Structures, Systems, and Components (SSCs) in a nuclear power plant need to be classified according to their importance to safety to protect people and the environment. The Specific Safety Guide, No. SSG-30, issued by IAEA, guides how to meet the requirements for identifying safety functions and classification of SSCs and ensuring their appropriate quality and reliability. However, the relevant guidance has not specified the perspectives of low radiological consequences on on-site workers. To refine the SSCs classification and complement the IAEA guidance, a proposed limiting value and corresponding assessment method for low radiological consequence caused by SSCs failure from the on-site workers’ perspective based on international good practices are developed, which include the main process of on-site radiological consequence evaluation, the calculation method for different exposure pathways and the assessment assumptions. Also, typical examples of SSCs failure are also assessed to validate the proposed method’s feasibility. The assessment results show that the activity concentration of failure SSCs, the workers’ locations, the exposure duration of workers, and the local dose rate of each location are the main factors for on-site radiological consequence evaluation, and the proposed method in this paper can assist in the safety classification of SSCs in the third generation of pressurized water reactor nuclear power plants.
Data Reconstruction of Faulty Sensors for the Nuclear Power Plants Control System: A Strong Tracking Filter Approach
Authors: Zhou Hongkuan , Zheng Wei , Tao Mo , Guo Xiaojie , Huang Chonghai , Chai Wenting , Chen Kai , Chen Zhaoxu 
Published: November 2022

Abstract
As an indispensable component of the control system, the sensor usually works in high temperature, high pressure, strong vibration and other failure-prone environment. The complex, changeable, and high intensity working environment makes the sensors work unstable, which eventually leads to the sensor fault. Once the sensor fault occurs, the output signal will deviate from the normal value, which seriously decreases the accuracy, stability and reliability of the control system. To solve this problem, the researchers tried to locate the faulty sensors and recover the missing data by the expertise, which results in low accuracy and long-time consuming. In this paper, a novel data reconstruction method for padding the missing data of the faulty sensor is proposed. Based on the extensive redundancy and complementarity of the information obtained from various types of sensors, the data reconstruction of the target sensor is realized by combining the thermal system model with the strong tracking filter algorithm. Simulation data of the single pressure condenser system show that, under different sensor fault modes, the reconstruction accuracy deviation of the signal reconstructed by the algorithm is less than 5% compared with the normal signal. The application of the proposed algorithm can significantly improve the safety and reliability of the control system.
Source Localization of Accidental Radionuclide Release Using a Data Assimilation Method Based on Forward and Backward Dispersion Simulations
Authors: Xu Yuhan , Fang Sheng , Dong Xinwen , Zhuang Shuhan 
Published: November 2022

Abstract
Source localization of accidental radionuclide release is crucial to nuclear emergency and decision making, especially when no reliable information about possible release regions can be previously obtained. To rapidly and accurately determine the source location with a limited number of measurements, this paper proposes a two-step data assimilation method based on forward and backward dispersion simulations in the RIMPUFF model. In the first step, meteorological variables such as wind speed are reversed and the measurements are set as sources for several backward RIMPUFF runs. By identifying the overlap of backward plumes originating from different measurement sites, potential regions discretized into grids can be roughly determined. In the second step, the grids selected by the first step are set as sources for forward RIMPUFF runs with unit release rate and the source location will be refined by minimizing a correlation-based function of simulated and measured data. The method is verified by SCK-CEN 41Ar field experiment. Using gamma dose rate data from all measurement sites, source location is retrieved with errors of 22.36m and 30m respectively on the two days of the field experiment. Compared with the direct correlation-based method, the proposed method achieves satisfactory solutions in an obviously shorter time. In particular, there is no need for predetermining the source release rate throughout the process, indicating that the method can conveniently combine with other source term inversion approaches whose source location is viewed as a known quantity. Therefore, the presented two-step data assimilation method here is potentially a useful tool with high accuracy and efficiency for further integration in nuclear emergency response system.
Validation of NUIT for Nuclides Composition Calculation With PWR Spent Fuel Isotropic Measurement
Authors: Tan Kai , Li Ruihan , Li Jian , Liang Jingang , She Ding 
Published: November 2022

Abstract
The NUIT (Nuclide Inventory Tool) is a reactor core nuclide inventory calculation code, which has been developed by Institute of Nuclear and New Energy Technology (INET), Tsinghua University. NUIT is able to compute instantaneous and integral values of various physical parameters including isotope composition, decay heat, particle spectrum, and radioactivity in fuel burnup, material activation, and decay calculations. The validation of the NUIT is an important process to ensure the correctness and applicability of the newly developed code. Presently, there is a lack of validation of the nuclide density calculation function of NUIT. Hence, this work is to validate this function of the NUIT code based on the PWR spent fuel isotopic compositions benchmark issued by the Oak Ridge National Laboratory in the SCALE program. In this work, the 103 spent fuel samples from 9 PWR plants are used to validate the nuclide composition calculation function of the NUIT, in which the initial enrichment of U-235 ranges from 2.453% to 4.657%, and the burnup ranges from 7.8 GWd/MTU to 60.7 GWd/MTU. The results show that NUIT’s nuclide density calculation results are in good agreement with the experimental values in general and the average error is about 10% for all nuclide density calculation cases. The typical nuclides, such as U-235, U-238, and Pu-239, etc., have an average calculation error of less than 5%, and most of the average errors of other actinide nuclides and fission products, such as Am, Cm, and Pu, etc., are within 10%∼40%, which meets the requirements of the radioactive source term analysis for the reactor core. Meanwhile, the comparison results show that the calculation accuracy of NUIT is comparable to other radioactive source term programs such as SCALE and OpenMC. The validation results further demonstrate NUIT’s capabilities in nuclide composition calculations.
RISMC Analysis Method Used in Operation and Maintenance of Nuclear Power Plant
Authors: Wang Zhao , Liu Weidong , Liu Leilei , Wang Yongxiang , Xu Liangwang , Pei Liang , Li Qiongzhe , Guo Jianbing 
Published: November 2022

Abstract
Risk informed safety margin characteristics (RISMC) analysis Method, is a safety analysis method which combines probability safety analysis and determinism safety analysis of nuclear power plant. General RISMC analysis platform “CARS” (Coupling Platform for Advanced Risk Simulation) can provide parameter distribution and Monte Carlo sampling function, and interactive coupling function with data of the thermal hydraulic safety analysis system program. The paper uses CARS and thermal hydraulic safety analysis system program to analyzes the technical requirements of personnel training and engineering modification of nuclear power plant, by means of accident scenario sampling, physical mechanism calculation and scenario consequence statistics. In the analysis process, various important factors which may affect the event process and consequences are comprehensively considered through the analysis of nuclear power plant operation rules, equipment reliability and the uncertainty distribution characteristics of key factors. And consequences of event scenario development, nuclear power plant response and the time course are analyzed through establish a realistic physical model by using thermal hydraulic safety analysis system program. The calculation results give the target level of personnel training and the effectiveness of engineering modification. The results show that RISMC analysis Method is an effective safty analysis and decision making method for operation and maintenance of nuclear power plants.
Reliability Analysis of Active Reactor Pit Flooding Function of EHR DFT System of a NPP Using an Enhanced Component-Connection Method
Authors: Guo Dingqing , Wang Shaoxuan , Ge Daochuan , Tong Lili , Cao Xuewu 
Published: November 2022

Abstract
Redundancy design is often used to enhance the reliability of complex systems which often run as safety-critical systems of nuclear power plant (NPP). Due to the redundancy design and management strategies, safety-critical systems of a NPP often have complex sequential failure behaviors. For a safety-critical system of a NPP with sequential failure behaviors, it is very important to evaluate its reliability for a given mission time. In this paper, the active reactor pit flooding function of EHR (containment heat removal system) of a nuclear power plant (NPP) is modeled by dynamic fault tree (DFT) and analyzed by BDD. To build the BDD, an enhanced component-connection-based method is proposed. The failure probability of the active reactor pit flooding function is analyzed. Besides, the FV, RAW and RRW importance of the EHR equipment are also evaluated. The results show the cumulative failure probability at mission time 24 hour is 1.81E−05, which is very low. The cumulative failure probability of train A (train B) is 4.25E−03. It can be concluded that the active reactor pit flooding function of EHR has a very high reliability given a mission time 24h.
Effect of Carbon Content on the Resistance to Intergranular Corrosion of Alloy 800H
Authors: Zhong Qiang , Wang Man , Sun Wen-Qiang , Wu Jun-Jie , Qian Jiong , Ning Sen-Sen 
Published: November 2022

Abstract
The effect of carbon content on the intergranular corrosion performance of Alloy 800H was studied by thermodynamic software simulation calculation combined with double-loop electrochemical potentiodynamic reactivation (DL-EPR), sulfuric acid-copper sulfate test, and microscopic morphology observation (optical microscope, scanning electron microscope). The results show that the main precipitation phases of Alloy 800H are α-Cr phase, M(C,N), M23C6, η phase and σ phase. At 500∼720°C, the higher the carbon content and the more M23C6 phase, the higher the susceptibility to intergranular corrosion. In the solid solution state, the Alloy 800H has low sensitivity to intergranular corrosion, and the carbon content has no obvious effect on its resistance to intergranular corrosion. In the sensitized state, carbides are continuously precipitated at the grain boundaries, and the intergranular corrosion susceptibility of the alloy increases rapidly.