bluuebunny/crossref_metadata_2025_split huggingface dataset:
Research on Sensing Value Reconstruction Method of Small Modular Reactor Based on Improved Short-Term and Long-Term Memory Neural Network
Authors: Cheng Shouyu , Zhu Hao , Zhang Bowen , Sun Jilin 
Published: November 2022

Abstract
Small modular reactor (SMR) is a complex, time-varying and strongly coupled nonlinear system, which adopts integrated design with short pipeline and small coolant inventory. Sensor failure of SMR under frequent load-following conditions threatens the safety of reactor. It is an effective way to realize sensor fault tolerance by abnormal sensing value reconstruction with machine learning method. In this paper, a sensing value reconstruction model combining hybrid dilated convolution neural network (CNN) and long short-term memory (LSTM) neural network is proposed. The hybrid dilated CNN is used to extract the feature information in the data set as the input of LSTM. LSTM is trained through training set to obtain the reconstructed sensor parameter time series. Besides, the traditional LSTM model is trained as a comparison model to verify the advantages of the reconstruction model. The training set is obtained from the SMR simulation model established by RELAP5 code. Multiple sensing values are input to each other’s reconstruction model. In order to avoid the one abnormal sensing value affecting the reconstruction behavior of other sensing value reconstruction models, the abnormal sensed value is replaced by its reconstructed value as the input of other models. The test set completely different from the training set is used to verify the behavior of the reconstruction model. The results show that the reconstruction model proposed in this paper not only fit the real data well, which means good reconstruction behavior on sensing value reconstruction, but also have better reconstruction effect compared with the traditional LSTM model.
Thermohydraulic Characteristics of Supercritical CO2 Flowing in Printed Circuit Heat Exchanger With Large Eddy Simulation
Authors: Jin Feng , Chen Deqi , Hu Lian 
Published: November 2022

Abstract
Printed circuit heat exchanger (PCHE) is a promising candidate of supercritical CO2 (S-CO2) Brayton cycle in the IV nuclear reactor with high efficiency, compactness and high resistance of temperature and pressure. In this paper, model of zigzag PCHE (D = 2.0 mm, Lp = 7.24 mm and θ = 100°) was simulated under the precooler condition near and away pseudo-critical region with S-CO2 and water as hot and cold fluid respectively. Method of large eddy simulation (LES) was applied to investigate the local performance of S-CO2 affected by both of complicated structure and drastic properties variation. The results showed that in the local position of a pitch, temperature and velocity on cross-section presented strong nonuniformity due to the disturbance of corner and symmetry on the side of corner, which showed heat transfer enhancement or weakness on different locations. Meanwhile, the turbulent structure of S-CO2 showed a periodic characteristic among pitches without influence by the drastic change of thermo-physical properties. Based on the results, two flow regions, convection region and mixed region were defined by dividing the local pitches. Combined with the vortex structure analysis using Q criterion, it was found that the mixed region presented more intensive vortex distribution and thus showed a higher heat transfer performance and flow resistance while convection region showed the opposite trend. Therefore, a fluctuation changing curve along the flow direction were observed on heat transfer efficiency and friction factor due to different characteristics of two regions.
Study on the Response of Prestressed Reinforced Concrete Slabs With High Reinforcement Ratio Under Low Speed Impact
Authors: Song Mengyan , Jiang Di , Li Changkai , Zhang Chaoqi , Zhi Xudong 
Published: November 2022

Abstract
Since the 9.11 incident, the impact of large commercial airliners on nuclear power plants and other important national facilities has attracted worldwide attention. The safety of flying objects hitting nuclear containment has become one of the important research topics in the field of nuclear power generation. In this paper, the impact resistance of non-prestressed and prestressed concrete slabs is studied through detailed experiments and numerical studies. By non-bonding rear sheeting, prestress of about 10MPa is applied to concrete slabs measuring 1400mm × 600mm, which are then hit by hemispherical steel hammers (590kg) at heights of 2m and 4m above the center of the span. The impact force, displacement, and the reinforced concrete slab and the prestressed concrete slab are compared. In LS-DYNA finite element software, the CSCM concrete structure model is used to study the numerical values of the experiments carried out. It has been found that with the increase of impact energy, the deformation produced by reinforced concrete slabs increases and is not recoverable. On the other hand, after applying presession, the impact force of prestressed concrete slabs was increased by 53.8% and displacement was reduced by 45% due to the increase of the stiffness of the plate. The maximum reproduction deviation of the impact force and displacement values is 17% and 14%, respectively. The research results will be applied to the structural dynamic impact protection design of nuclear power plant containment.
Interpretation and Analysis of Design Idea of the Spent Fuel Transport Ship
Authors: Zhong Gui , Hui Zhang , Wei Wang , Yueyan Yuan , Chao Liu , Xiaoqing Li 
Published: November 2022

Abstract
With the rapid growth of China’s nuclear power scale, in order to meet the future demand for spent fuel transport, relying on a single road transport mode is far from meeting the transport demand. It is necessary to build a spent fuel multimodal transport system to carry out large-scale spent fuel transport. As a special means of transport for sea transport in spent fuel multimodal transport, China’s first spent fuel transport ship is completely independently designed by China, has passed the empty ship trial voyage, and has also achieved good practice in actual transport. Based on the design requirements of domestic and international laws and regulations for the spent fuel transport ship, and combined with the actual situation of the wharf of domestic nuclear power plant, this paper interprets and analyzes the design ideas of the ship type size, loading layout scheme and safety system of the first spent fuel transport ship in China, in order to provide technical reference for the design and improvement of the spent fuel transport ship in the future.
New Opening Strategy of Filtered Containment Venting System
Authors: Zheng Hua , Wei Shuhong 
Published: November 2022

Abstract
Filtered containment venting system (FCVS) is used in some nuclear power plants to avoid containment overpressure failure and uncontrolled radioactive release in severe accidents. Currently the opening strategy is mainly based on the containment pressure, such as the design pressure. The containment has a stronger capability. There are various deposition mechanisms or even plugging of aerosol particles passing through potential micro channels within the containment. There are various natural or engineered deposition mechanisms of aerosol particles in the containment. The qualification pressure of FCVS can be increased. If the FCVS is opened later, the total radioactive release into the environment can be decreased, even though the leakage may be increased. A new opening strategy of FCVS is proposed, which is not just based on the containment pressure, but considers the containment capability, various deposition mechanisms or even plugging of aerosol particles passing through potential micro channels within the containment, various natural or engineered deposition mechanisms of aerosol particles in the containment. This strategy uses the radioactive release into the environment and associated radiological consequence as the optimization objective, considering both radioactive release through potential micro channels before venting and radioactive release after venting, and can reduce the radioactive released into the environment as far as possible while maintaining the integrity of the containment.
Research on the PID Parameter Self-Tuning of the Pressurizer Control System of Nuclear Power Plant
Authors: Xing Kunpeng , Wang Hongrui , Zou Tao , Yang Zhijia , Wang Jingyang , Zhang Pengbin 
Published: November 2022

Abstract
The pressurizer is an essential part of the safe operation of the nuclear power plant. It contains both the steam phase and the liquid phase, which requires simultaneous control of the pressure and water level, and the control system is strongly coupled. When using the PID control method, it is difficult to tune the parameters, and the difficulty is exacerbated when the set value fluctuates. This paper aims at the multi-input and multi-output pressurizer control system, based on the particle swarm optimization to realize the PID controller parameters self-tuning, adopting the time and absolute error integral as the fitness function to evaluate the merits of the PID parameters, and constantly iterating to update and obtain the PID parameters that better satisfy the current working conditions. For coupled pressurizer control systems, the PID parameter tuning method using particle swarm optimization algorithms can give a group of PID controller parameters that make the system operation stable despite the frequent fluctuation of the set value, and compared with the traditional PID controller parameter tuning method, this method can realize the PID parameter self-tuning, eliminating the trouble of manual tuning.
Thermal-Hydraulic Analysis of HL-2M Cooling Water System for Toroidal Field Coils
Authors: Wang Jiaqi , Xie Yanfeng , Liu Xiaolong , Meng Jianpeng , Jia Ruibao , Li Qiang , Chen Xin 
Published: November 2022

Abstract
HL-2M is a medium-sized copper-conductor tokamak from the Southwestern Institute of Physics (SWIP). HL-2M cooling water system is composed of Tokamak Cooling Water System (TCWS), Component Cooling Water System (CCWS), and the Heat Rejection System (HRS). The tokamak cooling water system, the primary cooling system of the tokamak machine, is further divided into TCWS-1, TCWS-2, TCWS-3, TCWS-4 depending on the water chemistry needs of clients and pressure requirements. As an important subsystem, the function of the TF coils cooling water system (TCWS-1) is to remove heat load from Toroidal Field coils. HL-2M copper TF coil is made up of 20 coil bundles, each consisting of seven turns for 140 turns, each of turns with six cooling channels for 840 cooling channels. A thermal-hydraulic model of TCWS-1 magnet coil client was established, by COMSOL Multiphysics, to conduct the transient state thermal-hydraulic analysis. The TCWS-1 cooling loop was developed, by using AFT Fathom code, to conduct the steady-state thermal-hydraulic analysis of the system. In this thermal-hydraulic analysis model, by multi-physical coupling calculations on the client and applying the results to the TCWS-1 loop calculation, more realistic results can be obtained. This paper presents the results of this thermal-hydraulic analysis which was composed of the main thermal-hydraulic parameters for TCWS-1 loop and client, including flow rate, velocities, pressure drops, and outlet temperatures. The results obtained in this study can provide support and reference to the engineering design of TCWS-1.
Study on Separation of Palladium From High Level Liquid Waste by Potassium Manganese Ferrocyanide
Authors: Qi Tianming , Sang Hongji , Mao Cong , Wen Yueying , Hu Jipu , Wu Yan 
Published: November 2022

Abstract
Glass solidification is a favorable treatment method for high level liquid waste (HLLW) from spent fuel reprocessing in industrial application, and the presence of palladium (Pd) in HLLW may seriously affect the glass solidification process. Ferrocyanide with a perovskite-like face-centered cubic structure has strong adsorption affinity towards Pd. In this study, silica-based composite KMnFC/SiO2 was prepared by pore crystallization of potassium manganese ferrocyanide (KMnFC) into porous SiO2. A series of characterization of the synthesized adsorbent were carried out such as XRD, TG-DTA, and SEM-EDS. It was verified that the adsorbent was successfully prepared and it has good Microscopic structure along with element distribution.
A series of adsorption experiments were carried out. The adsorption of Pd is not easily affected by the concentration of nitric acid, which shows good acid resistance. Through the adsorption isotherm and adsorption kinetic curve, it is proved that the adsorption process of Pd by KMnFC/SiO2 is chemical monolayer adsorption. The adsorbent has strong adsorption selectivity for Pd in multi-ionic solution. Under the condition of complete adsorption of Pd, the adsorption rate of the adsorbent for other elements except Ru is less than 20%.
Improvements on Fuel Failures Diagnosis Method for Pressurized Water Reactor Nuclear Power Plant Based on Radioactive Activities Inside the Waste Gas Treatment System
Authors: Lyu Weifeng , Guo Runchun , Ran Wenwang , Gao Yaoyi , Xiong Jun 
Published: November 2022

Abstract
The cladding of fuel elements in the core of a pressurized water reactor nuclear power plant (PWR) is the first barrier to the radioactive material in the reactor. Once the fuel cladding fails, the radioactive fission products or even fissile material in the fuel elements will be released into the primary coolant, thereby challenging the safe operation of the plant. Fuel cladding failures diagnosis is indispensable for PWR as fuel failure is not inevitable and the impact on nuclear safety is significant. Based on the continuous research on fission product generation and release mechanisms for decades, several fuel failure diagnosis methods for PWR have been established and specific nuclides have been selected as indicators of fuel failures. Recent research shows Kr-85 is a good indicator for the diagnosis of the number of failed fuel rods for its insensitivity to the fuel rod defect size and failed fuel burn-up, but it is hard to measure the activity of Kr-85 in the primary coolant due to the interference of other nuclides. However, the activity of Kr-85 inside the waste gas treatment system can be measured where the activity of other nuclides has been significantly reduced by decay. Hence, the fuel failure diagnosis method is improved based on Kr-85 activities inside the waste gas treatment system. The operating data of Chinese in-service nuclear power plants are used to validate the improved fuel failure diagnosis method, and the validation results show that the improved diagnosis method of fuel failures can accurately diagnose the fuel failure and has wider applicability.
Research and Development of Default Value Analyses and Management Platform for Nuclear Power Plant Control System
Authors: Zhao Hongbin , Liu Zhiyong , Liu Hemin , Liu Weifang 
Published: November 2022

Abstract
The research and development of default value development and management platform of nuclear power plant control system can realize the functions of default value analysis, value selection, query and data management. The software standardizes the process and judgment factors of default value analysis process, makes qualitative and quantitative judgment on value selection factors, and combines the value selection algorithm embedded in the software, The auxiliary designer can automatically get the default value. The use of software can effectively improve the rationality and standardization of default value analysis. The software platform solves the problems of different default value analysis standards and chaotic data management in nuclear power plants. As the first special software for signal default value analysis and management of nuclear power control system in China, it has improved the intelligent level of the third generation nuclear power HPR1000.
Research on Measurement Method of a Device for Measuring the Perpendicularity of Tubesheet Hole
Authors: Xu Xianghua , Hu Guihong 
Published: November 2022

Abstract
The tubesheet is the key part of the steam generator, and the quality of the tubesheet hole is the key characteristic of the steam generator. The inspection of tubesheet holes includes defects on the surface of the tubesheet holes and the geometric dimensions of the tubesheet holes. The geometric dimensions of the tubesheet hole mainly include hole diameter, perpendicularity, hole bridge, position and roughness, etc. The perpendicularity of tubesheet hole is an important characteristic, which directly affects the size of tubesheet secondary hole bridge and heat exchange tube installation. This paper describes a device for measuring the perpendicularity of tubesheet hole, discusses the principle and measurement method of the tubesheet hole perpendicularity measuring device, and analyzes the error and proposes measures to improve the measurement accuracy of the device. Analysis of data processing provides a basis for the measurement device and measurement method to meet the requirements of specifications. The results show that the designed measuring device to measure perpendicularity has the advantages of simple operation, simple measurement and high accuracy of measurement results. It is beneficial to ensure the quality of heat exchange tube installation of the steam generator, which can speed up the manufacturing progress of the steam generator.
Practical Issue Analysis of Miniature Circuit Breaker Implementation
Authors: Xingwei Fei , Kunpeng Tian , Yu Zhao , Ping Li 
Published: November 2022

Abstract
Miniature circuit breakers are widely used in engineering applications. This paper analysis the operation principle of overload protection and short circuit protection from its structure. Based on the actual situation in unit 3 and 4 of Qinshan Nuclear Power Plant, the basic reasons for several typical problems which have occurred on the miniature circuit breaker are analyzed and the treatment scheme is given respectively. Aiming at the defect that the breaker incorrect triped in main transformer cooling control-box in the high temperature environment, combined with the temperature coefficient correction and the parallel arrangement of the drop capacity effect, the miniature circuit breaker is rechosen to avoid the defect. In view of the defect that the upstream breaker of the PAMS cannot be switched on during the surge current, by giving the analysis of the match between the electromagnetic release characteristics of the miniature circuit breaker and the impact current of the downstream parallel load closing, the scheme is put forward to increase downstream parallel load breakers. According to the selectivity of the miniature circuit breaker, the unreasonable problem of the last stage distribution in the protection configuration is found out in view of the problem of RPR protecting the internal short circuit of the cabinet, and according to the requirements of the national standard, the selective test verification of the related model miniature circuit breaker is carried out. The reasonable part has been changed and carried out on the spot. After the actual operation of the unit, three typical problems occurred on the miniature circuit breakers have been effectively solved.
Design Improvement of the Discharge Proposal Preventing SG Overfill After SGTR in HPR1000 PWR
Authors: Li Qinghua , Fang Jun , Li Yafei , Zhao Bin , Wang Guangfei 
Published: November 2022

Abstract
Base on the result of full scope accident analysis for HPR1000,which is a third generation localized PWR with combination of initiative and passivity, the secondary side pressure of the damaged SG should be controlled by the steam generator blowdown system (TTB) during the retreat process of SGTR accident. In order to response the safety criteria of DBA function, the proposal of post accident discharge pipeline of TTB system should be improved. The part of secondary side fluid are discharged through the pipeline after SGTR, to realize the decrease of water level of damaged SG and preventing SG overfill. The paper suggests a proposal of post accident discharge pipeline for TTB as a mitigation measure of SGTR accident. The discharge flowrate with two phases flow condition in the discharge pipeline must be verified by theoretical calculation to evaluate the mitigation capability of the proposal. Now the performance test has been finished in the first unit of HPR1000, The achievement indicate the discharge capability of the proposal conform to the relative safety criteria for the SGTR accident.
General Safety Assessment of Seismically Induced Hazards for NPPs
Authors: Yang Guanghua , Yang Cuilian 
Published: November 2022

Abstract
Design basis earthquake may cause various internal hazards in nuclear power plants, such as fire, explosion, flooding, missiles, high-energy pipelines break, etc. In addition, earthquakes may also cause non-seismic items to fall or collapse, thus affecting safety-related systems or equipment. As the subsequent situation induced by earthquake is quite complex, no practicable methods have been presented yet to be used for the safety evaluation after earthquake. This paper introduces a feasible safety assessment concept for the above-mentioned hazards caused by the design basis earthquake based on the engineering practice experience.
This general assessment concept based on the deterministic approach could be used for guidance of safety evaluation during or after earthquake for nuclear power plants.
Research on Development of On-Line Monitoring of Fuel Damage in Nuclear Power Plant
Authors: Yiduo Tian , Wenjun Hu 
Published: November 2022

Abstract
The occurrence of fuel damage has serious impact on safety and economic efficiency of nuclear power plants. The on-line monitoring and diagnosis system of fuel damage in nuclear power plant can timely determine whether the fuel element cladding is damaged or not, and diagnose the relevant information such as the size of the rupture, the position of the damaged component and the location of the damage.
The paper investigates some fuel damage monitoring systems and analyses their basic structure and process. Based on the examples, the paper classifies the process into radioactive signal detection, decomposition and failure diagnosis part and discusses the theoretical and experimental basis. Different methods of fuel damage diagnosis along with factors resulting in errors and applicable conditions are also discussed.
Through the investigation, some results can be concluded that the reason that the online detection system of fuel damage is not widely used is mainly due to precision limitation. Future efforts should be made on decomposition and failure diagnosis part in order to reach a fast response and high accuracy.