bluuebunny/crossref_metadata_2025_split huggingface dataset:
The Effect of Nozzle Safe End Hot/Cold Wire Tungsten Inert Gas Buttering on Microstructure and Property
Authors: Yang Chengdong , Zhang Maolong , Ru Xiangkun , Tang Weibao 
Published: November 2022

Abstract
Hot wire tungsten inert gas (HWTIG) and cold wire tungsten inert gas (CWTIG) have been used for nozzle safe-end buttering of nuclear island critical equipment, the effect of the HWTIG and CWTIG buttering on microstructure and property has been investigated. Research results revealed that the grain orientation of HWTIG buttering layer is mostly concentrated in (0 0 1) direction and the average grain size is 118.375μm, while the grain orientation of the CWTIG dispersed distribution in (0 0 1), (1 1 1) and (1 0 1) direction, and the grain size is 158.479μm. Furthermore, the grain boundary misorientation angle of HWTIG buttering layer has higher occurrence frequency between 10° and 30° high-angle boundary, whereas the CWTIG has higher occurrence frequency within 30° and 60°. Moreover, the microstructure of HWTIG buttering layer has less chromium-rich M23C6-type carbide compared to CWTIG, and the M23C6-type carbides discontinuous distribution on the grain boundary of HWTIG buttering layer. The buttering layer of HWTIG has lower ductility-dip cracking sensitivity than CWTIG.
Level 3 PSA Application in a UK Generic Design Assessment Project
Authors: Wang Jinkai , Wang Qi 
Published: November 2022

Abstract
Level 3 Probabilistic Safety Assessment (L3 PSA) is required in UK Generic Design Assessment (GDA) to demonstrate that a new nuclear power plant is suitable to be built in UK. L3 PSA is used to assess the individual and societal risk and compare the results against the offsite Radiation Protection Targets (RPTs) for fault and accident conditions. There is little relevant good practice for L3 PSAs that have recently been performed worldwide and there is lack of mature standard to implement L3 PSA. In this study, a pilot L3 PSA is performed for UK HPR1000 to reflect the UK context and relevant good practices. It introduces the methodology and the processes to be followed to perform conditional consequences calculations for the faults and accident scenarios. All radiation sources are considered and analyzed. The radiological risks to a potential UK site are analyzed and compared against RPTs. A widely used code - PC COSYMA, is selected as a primary tool for the accident consequence calculations. The strengths and limitations of the code are identified based on the project situation, and either qualitative arguments or supplementary analysis is subsequently proposed to overcome the limitations. The final L3 PSA results has supported the demonstration that the offsite radiological risks for UK HPR1000 have been achieved as low as reasonably practicable (ALARP) and meet the UK regulatory expectation.
Effect of Cr6+ Concentrations on the Corrosion Behavior of 316LN Stainless Steel in Supercritical Water
Authors: Li Zhenhe , Huang Xi , Zhan Zixiong , Zhao Qi , Liu Junxiong , Wei Lihua , Li Xiaoyan 
Published: November 2022

Abstract
It is well known that the type and concentration of impurity ions in pressurized water reactor environments will significantly alter the corrosion behavior of materials. However, the effect of these water chemistry parameters on the corrosion behavior of candidate materials for future supercritical water-cooled reactor (SCWR) systems is still a lack of understanding. The type and concentration of impurity ions on the corrosion behavior of candidate materials for future supercritical water-cooled reactor (SCWR) systems is still a lack of understanding. In this study, the effect of Cr6+ concentrations on the corrosion behavior of 316LN stainless steel in 550 °C/25 MPa supercritical water (SCW) was investigated. The results show that the weight gain first decreased and subsequently increased with the increasing of Cr6+ concentration, which is 0.40±0.01 mg/cm2 for 0 ppm, 0.28±0.03 mg/cm2 for 0.1 ppm, 0.55±0.02 mg/cm2 for 1.0 ppm, and 0.59±0.01 mg/cm2 for 50 ppm, respectively. The Fe-rich magnetite oxides were gradually covered on the surface of samples, and its proportion on the surface followed a similar alteration trend as the weight gain, which is 44.3% for 0 ppm, 37.8% for 0.1 ppm, 67.7% for 1 ppm, and 87.6% for 50 ppm. A duplex structure of the oxide films, including an Fe-rich outer magnetite layer and a Cr-rich inner spinel layer, was developed on the regions covered with Fe-rich magnetite oxides. The alternation in the thickness of oxide film followed a similar trend as the weight gain, which is 7.67 μm for 0 ppm, 5.80 μm for 0.1 ppm, 9.97 μm for 1 ppm, and 11.23 μm for 50 ppm. However, a Cr-rich spinel oxide film was formed on the region without Ferich magnetite oxides. The thickness of this oxide film was unchanged regardless of Cr6+ ion concentration, which is about 2 μm, but the concentration of Cr in this oxide film is significantly higher than reference sample. These results indicate that the increasing of Cr6+ concentration in SCW will result in the corrosion rate to decrease first and then increase, which means that some measurements should be conducted during the operation of SCWR to keep a lower concentration of Cr6+.
Research on Source Characteristics of Radionuclide Diffusion in the Ocean Under Oceanic Radioactive Events
Authors: Li Zichao , Chen Rongchang , Liu Chen , Xue Qingqing , Wang Zhixia , Zhou Tao 
Published: November 2022

Abstract
All operational and under construction nuclear power plants in China are distributed in coastal areas presently. Facing the problem of spent fuel from nuclear reactors, we are also actively promoting the system construction of combined transport with highway, ocean and railway. In case of a nuclear leakage accident, radionuclides will diffuse through the atmosphere and the ocean, and pollute the ecological environment. To solve the problem of radionuclide diffusion in the ocean, we first need to solve the source term problem. The study on source characteristics of radionuclide diffusion in the ocean under oceanic radioactive events plays an important role in simulation and emergency decision-making. The meaning of source term and unit parameters were given, and types, forms and release modes of radionuclides were analyzed under nuclear leakage accidents. The storage and release of radionuclides were calculated. Then the dry and wet deposition fluxes of radionuclides were calculated. The results show that 131I, 134Cs and 137Cs are the most important radionuclides in radiation dose evaluation and they will be in the form of CSI and CsOH under nuclear leakage accidents. Radionuclides are released into the environment in the form of micron aerosol. Radionuclides entering the ocean include direct release and atmospheric radionuclide deposition. The dry deposition flux is 3 orders to 4 orders of magnitude lower than the ground radionuclide concentration. The wet deposition flux within 300m is much larger than the dry deposition flux, and the wet deposition flux outside 300m is about 1 order to 2 orders of magnitude larger than the dry deposition flux. It provides a prerequisite for the calculation of radionuclide diffusion in the ocean under oceanic radioactive events.
Research on Radionuclide Diffusion and Verification in the Ocean Under Nuclear Leakage Accidents
Authors: Li Zichao , Chen Rongchang , Xue Qingqing , Liu Chen , Wang Zhixia , Zhou Tao 
Published: November 2022

Abstract
All operational and under construction nuclear power plants in China are distributed in coastal areas presently. Facing the problem of spent fuel from nuclear reactors, we are also actively promoting the system construction of combined transport with highway, ocean and railway. In case of nuclear leakage accidents, radionuclides will diffuse through the ocean and pollute the ecological environment. The study of radionuclide diffusion and verification in the ocean plays an important role under nuclear leakage accidents. The hydrodynamic model and radionuclide diffusion model in coastal waters of nuclear power plant were established. The water level change and radionuclide diffusion area in coastal waters of nuclear power plant were calculated, and dangerous areas were given. The calculated results were verified by enteromorpha drift. The results show that radioactivity in the high concentration area can reach 109Bq/m3, and it is reduced by about 9 orders of magnitude to the source term. On the 14th day, the dangerous area is within about 58km to the east, 10km to the south and 21km to the west of the nuclear power plant. In the east-west direction, the dangerous area is about half of the radionuclide diffusion area.
Nuclear Power Plant Design Optimization Based on Risk-Informed Method
Authors: Li Wenjing , Liu Yu , Wang Gaopeng , Yu Xinli 
Published: November 2022

Abstract
The probabilistic safety analysis method is an effective tool to demonstrate the balance of design and identify the vulnerabilities. It can provide technical support for the design optimization of nuclear power plant. A nuclear power plant in China specially designed the reactor cavity injection system (RCI) with both active and passive trains to mitigate the severe accident with core damaged. When the core damage accident occurs, the melt debris in the lower head can be cooled by RCI with water injection to the outside surface of the reactor vessel, thereby maintaining the integrity of the vessel and realizing the core melt in-vessel retention (IVR). This paper will introduce the design of the cavity injection system, and use the PSA method to establish a model of the system. Then the reliability of the system and its contribution to the large release frequency will be discussed, the vulnerability of the system will be identified. System design optimization recommendations will be proposed based on PSA results, and the results comparison between the optimization scheme and the original design is discussed. It can be concluded that in the design process of NPP, risk-informed methods should be considered as an important technical support tool.
Flow Distribution Optimization Design of Certain Integrated Pool Lead-Bismuth Reactor
Authors: Wang Na , Zhou Youxin , Yuan Zhaojun , Liao Ziyu 
Published: November 2022

Abstract
The core inlet flow distribution of core is of great importance in the integrated pool lead-bismuth reactor design, while the integrated pool structure feature and lead-bismuth fluid bring difficulties to the flow distribution optimization. The result of flow distribution will directly affect the safety margin of thermal-hydraulic design: mal-distribution could lead to decrease of core thermal safety margin, which may cause negative effect on the reactor safe operation. Based on three-dimensional computational fluid dynamics (CFD) simulation, the flow distribution result is evaluated. It’s found that the eddy flow at the core entrance will lead to nonuniformity of flow distribution and flow fluctuation. Optimizations of design structure are conducted to improve the flow distribution characteristic. Different options of flow channel and flow distribution device design are proposed. By means of steady and transient calculation, the optimization options are evaluated from the two aspects: the uniformity and the stability of flow distribution. The local resistance loss and flow velocity are considered at the same time. At the end of the research, the structural design scheme which has better flow distribution characteristic and meets the restriction of local resistance loss and flow velocity is put forward. The research could provide reference to the thermal-hydraulic design of integrated pool lead-bismuth reactor.
Seismic Response Analysis of Pile Foundation of an Underground Storage Facility of Nuclear Power Project
Authors: Wang Dongyang , Sun Xiaoying , Liu Ziqiao , Gan Yingying 
Published: November 2022

Abstract
The proposed site condition of an underground storage facility of nuclear power project (NPP) is complex, characterized by strong spatial variability of bedrock surface, non-uniform soil layers and underlying soft soil. If the substructure method commonly used in nuclear power engineering seismic analysis is adopted, the non-homogeneous foundation must be simplified as the horizontal layered site condition. Therefore, the dynamic interaction features among piles, foundation soil and superstructures under strong earthquakes cannot be reflected accurately. Taking it as the project background, this paper conducts a research on the three-dimensional time-history seismic analysis method for NPPs. On the theoretical basis of viscoelastic artificial boundary and wave equation method, the soil-structure dynamic interaction analysis methodology in complicated non-homogeneous site is put forward in this paper, which is called the integral time-domain method. And an efficient platform is set up for computing artificial boundary parameters and seismic motions. Then the seismic response analysis of pile foundation of the underground storage facility is completed. It is concluded that the pile-soil-structure interaction does not affect horizontal response of the super structure significantly but leads to a frequency shift of the response peak in vertical direction. Additionally, the substructure method is used to verify the accuracy of the proposed method and the results of two methods match well. In conclusion, the integral time-domain method contributes to appropriately simulating the foundation radiation damping and seismic input. It has high accuracy and applicability. The research method and the computing platform can be taken as technical references for seismic response analysis of NPPs constructed in non-horizontal layered foundation or complex geologic bodies conditions.
Research on a Circuit Health Assessment Technology Combining Mechanism Knowledge and Data Driven Technology
Authors: Shi Guilian , Xiaojin Huang , Changyu Mo 
Published: November 2022

Abstract
Degradation of electronic equipment is typically accompanied by a deviation in their electrical parameters from their design values. When the function of the electronic equipment fails, parametric faults of the electrical parameters will ultimately occur. Such electrical parameters can characterize the state of health can be defined as health parameters. For complex electronic equipment such as hardware circuits of Nuclear power plant DCS cards, the use of health parameters for fault diagnostic has been widely used, but there is no mature method for sub-health diagnostic. To address this problem, an electronic equipment health assessment method was developed using a technology based on the combination of mechanism knowledge and data-driven. The method first uses Physics of Failure approach and hardware circuit logic principle to identify the major failure mechanisms of the circuit, and then determines the key health parameters and their sub-health thresholds on this basis. The time series data under specific test conditions for the key health parameters of the circuits to be assessed is collected and compared with the sub-health thresholds, then the health status assessment results based on mechanism knowledge can be obtained. As well, the data-driven health assessment technology based on Mahalanobis distance is applied to analyze the real-time health parameter data of all samples, and the healthy outlier samples are identified and quantified, and the health status assessment results based on the data-driven technology can be obtained. The results of the two are comprehensively evaluated to finally determine the circuit health status. Practice was carried out on the DCDC chip of a nuclear power DCS analog output module, and the results showed that this method was feasible for engineering.
HL1000 Nuclear Power Station Emergency Borating System Capacity Demonstration
Authors: Tian Zhaohua , Li Qiang 
Published: November 2022

Abstract
Emergency Borating System has been widely used in the new third generation nuclear power plant in the world. HL1000 nuclear power plant is equipped with the Emergency Borating System (RBS) as an engineered safety facility, which provides the reactor the high pressure safety boron injection and long-term boration functions to ensure the core in subcritical. Firstly, the main functional requirements of RBS system of new third generation nuclear power plants in the world are introduced. Then, according to the two functional requirements for RBS of HL1000 nuclear power plant, reactivity control and confinements, the design criteria for RBS are determined. Based on these design criteria, the methodologies for RBS capacity demonstration are developed by the multi status point method. The relevant capacity parameters of RBS, including the maximum and minimum injection flow of RBS and the minimum volume requirements of boric acid tank are calculated separately.
Study on Interfacing System LOCA of VVER-1200
Authors: Du Jinyan , Dong Fangyu , Ma Chao 
Published: November 2022

Abstract
The evaluation of public risk at nuclear power plants has generally found the risk to be very low. Part of the reason for the low assessed risk can be attributed to the “defense-in-depth” philosophy of nuclear power plants. This defense-in-depth includes multiple barriers to prevent the release of radionuclides to the environment. Nevertheless, accident scenarios have been postulated which might compromise the defense-in-depth by bypassing the reactor coolant system boundary and the containment. One such accident type is referred to as the interfacing system LOCA (ISLOCA). This type of sequence involves the loss of isolation between the high pressure RCS and a low pressure system outside containment. ISLOCA could lead to the loss of coolant outside containment while simultaneously disabling ECCS injection. Such a scenario could result in core damage and might have substantially higher consequences than some of other postulated severe accidents because the containment is also bypassed. Thus, it is necessary to calculate the ISLOCA risk for a specific nuclear power plant. Based on summarizing the ISLOCA analysis method and process, this article carried out the ISLOCA analysis for the VVER-1200 nuclear power plant, identified its typical ISLOCA path, and quantified the impact on the safety of nuclear power plants.
Automatic Scram Forecasting During Abnormal Conditions of Nuclear Power Plants Based on LSTM
Authors: Chen Hanying , He Xiaoqiang , Gao Puzhen , Tan Sichao , Hou Xiaofan 
Published: November 2022

Abstract
During abnormal conditions of nuclear power plants that do not lead to automatic scram immediately, operators need to monitor the transient condition of nuclear power plant in order to evaluate the development rate of the abnormal conditions. Conventional condition monitoring methods cannot provide operators with an alert or forecast for automatic scram. In this work, a deep-learning model is presented for predicting the automatic scram assuming no operator intervention. The presented prediction model is based on Long Short-Term Memory (LSTM) model, which is a special type of gated Recurrent Neural Network (RNN) designed to adaptive control the memory length of the learned features and showing good performance in Sequence to Sequence (Seq2Seq) problems. The predicting model is trained by transient data of abnormal conditions: the input of model is the nuclear power plant monitoring data; and the output is the remaining time from current time to reactor trip. The predicted remaining time to reactor trip decreases with the development of abnormal condition, thus the output of prediction model will generate a countdown to reactor trip. Prediction experiments were conducted on simulated abnormal condition data generated by a full-scale simulator for a 300MW pressurized water reactor (PWR). The data with artificial noise were also used to test the performance of the presented model. The experiment results shows that the presented prediction model has better prediction performance and is more robust to noisy input than models based on one-hidden-layer feedforward neural networks or simple RNNs (e.g., Elman network). The presented intelligent prediction model can provide nuclear power plant operators with a countdown alert about the upcoming reactor scram, as such the operators can prepare or take countermeasures for it.
Simulation on the Multidimensional Behavior of TRISO Particle Fuel With UN/ UO2 Kernel
Authors: Wu Changwei , Wu Junmei , He Yanan , Wu Yingwei 
Published: November 2022

Abstract
As one of the new advanced fuels proposed to improve the safety of nuclear reactors, fully ceramic microencapsulated (FCM) fuel consisting of TRistructural ISOtropic (TRISO) particles dispersed in a silicon carbide matrix has received a lot of attention at home and abroad because of its distinguished advantages such as high corrosion resistance and excellent fuel irradiation stability. For analyzing FCM fuel performance and optimizing FCM fuel design, it is necessary to assess complicated behaviors of TRISO particle fuel under irradiation. To investigate the multidimensional behavior of TRISO particle fuel with UN or UO2 kernel, a three-dimensional thermomechanical coupling model was developed based on multiphysics coupling finite element software COMSOL. Additionally, thermomechanical performance analysis of TRISO particle fuel with irradiation-dependent material properties and behavior models was conducted, and important phenomena were implemented such as fission gas release, production and diffusion of CO gas, as well as the evolution of particle radius due to the irradiation deformation, fission product swelling strain and irradiation creep. Moreover, on the basis of the preliminary thermomechanical analysis, the performance of TRISO particle fuel with UO2 and UN kernel were compared. This paper is intended as an exploratory investigation of TRISO fuel behavior, which will provide vital reference for further performance analysis and design optimization of subsequent FCM fuel.
Design and Commissioning of Digital Control System for Refueling Machine in HPR1000
Authors: He Zhijun , Zhang Peng 
Published: November 2022

Abstract
The design of digital control system for refueling machine in HPR1000 fully absorbs and refers to the experience of independent design and supply of refueling machine in the previous PWR projects, the experience of technical service in the operating nuclear power plants and the experience feedback from owners.
This paper describes the design scheme, improvement and commissioning of refueling machine digital control system for HPR1000 in detail. It focuses on the design of system architecture, hardware selection, PLC system configuration, PLC program development and HMI configuration, etc. At the same time, in order to reduce the work intensity of operators, the digital control system enhances the function of HMI, improves the display of fault information, and realizes the automatic digital control of refueling machine. In HPR1000 reactor, the digital control system of refueling machine has been optimized and improved in both hardware design and software design, which greatly improves the safety, reliability and economy of refueling machine.