bluuebunny/crossref_metadata_2025_split huggingface dataset:
Calculation of Detector Response Rate Based on BEAVRS Benchmark Using PWR-Core Analysis Code CORAL
Authors: Yang Wen , Liu Baolin , Yao Shiwei , Chao Fei , Li Xing , Qiu Jinrong 
Published: November 2022

Abstract
Three dimensional PWR-core analysis code CORAL developed by Wuhan Second Ship Design and Research Institute, which provides all functions required by PWR-core analysis calculation. These functions are neutron diffusion within the core and reflector, macroscopic depletion or microscopic depletion calculation analysis, multi-channel or sub-channel thermal-hydraulic analysis, one-dimensional heat transfer from nuclear fuel to the coolant, critical search by boron concentration or control rod position, integral and differential worth of neutron absorbers, neutron kinetics parameters for transient analysis, in-core neutron detector response simulation etc. CORAL is convenient to update and maintain in consider of modular, object-oriented programming technology. In order to verify the computational capabilities of the reactor core analysis code, the BEAVRS benchmark is adopted as the research object. The BEAVRS problem is a benchmark problem based on real commercial PWRs with detailed material description, geometric information, core operating history, and detector measurement data. In this paper, the CORAL code is used to carry out physical analysis and calculation for the BEAVRS benchmark, and the response rate distribution of radial and axial detectors can be obtained. By comparing with the measurement results provided by the benchmark, it can be found that the calculation results of the CORAL code are in good agreement. This shows that the CORAL code can well simulate the detector response distribution in the core.
Experimental Investigation on the Characteristics of Bubble Departure Frequency of Subcooled Flow Boiling in Narrow Rectangular Channel
Authors: Chen Yong , Liu Hanzhou , Chen Deqi , Liu Haidong 
Published: November 2022

Abstract
Subcooled flow boiling in narrow rectangular channel is a very effective heat transfer method and is very important to nuclear reactor design and thermal hydraulic safety. The departure frequency of bubbles in subcooled boiling flow has direct influence on heat transfer. Subcooled flow boiling experiment is carried out to investigate the effect of different working conditions on the bubble departure frequency in the narrow rectangular channel. Because the test section is composed of quartz glass and Sapphire crystal Glass, it can be observed from four different directions to obtain high quality images. The experiment is conducted with mass fluxes of 400 to 1500 kg/(m2s) and heat flow density between 50∼300 (KW/m2). Besides, the system pressure is 100 kPa to 700kpa and inlet subcooled is 10 to 60 °C. Two types of nucleation sites can be detected from the experimental results: sliding growth and condensation. The bubble departure frequency of sliding is significantly higher than that of condensation. Besides, the effect of different working conditions on the bubble departure frequency in narrow rectangular channels was investigated. Finally, the departure frequency is analyzed and compared with previous models. The results show that the model proposed by Cole is in good agreement with the experimental results.
Verification and Validation of JMCT Monte Carlo Code in CPR1000
Authors: Ye Yaoxin , Zhao Jun , Bao Pengfei 
Published: November 2022

Abstract
The Monte Carlo method is essential in field of nuclear engineering and has been widely applied in reactor particle simulation owing to the rapid development of computer technology. Academic institutes and organizations of different countries have developed their own Monte Carlo code. The JMCT code is a Monte Carlo particle transport code developed by the China Academy of Engineering Physics. Before the JMCT Monte Carlo code is applied to the design of new nuclear power plant, the verification and validation (V&V) of the code on the operating reactors is necessary. In this work, the V&V of the JMCT code on criticality calculation and fixed-source calculation was performed on the operating CPR1000 reactors. First, the full core pin-by-pin model of CPR1000 was established and the criticality calculation was performed. Then, the 1/4 core model with irradiation surveillance capsules (ISC) of CPR1000 was established and the fixed-source calculation was performed. The calculation results of JMCT were compared to the reference code and the measured values in the ISC.
Comparative Model of Nuclear Power Plant General Layout Scheme
Authors: Huang Chong Guang , Wu Yan 
Published: November 2022

Abstract
Nuclear power plant sites are scarce resources for nuclear energy development. It is necessary to optimize the general layout scheme of nuclear power plant to reduce the project cost and operation cost. In the feasibility study stage of the project, the determination of the general layout scheme of nuclear power plant can provide design input for the land acquisition negotiation, geological prospecting, environmental assessment, site safety analysis and other topics. Then the rationality of the nuclear power plant project is proved more strongly.
The work content of general drawing runs through the whole process of power plant engineering construction, involving many aspects such as seismic geology, hydrometeorology, and bulky transportation. Among them, the comparison and selection of the general layout scheme of nuclear power plant is a key step of the demonstration in the early stage of the project, and it needs to be iterated with several topics in the early stage to be reasonably determined. Therefore, it is very important to find a method to compare and choose the optimal general layout scheme of power plant.
In this paper, a fuzzy comprehensive evaluation method based on AHP is proposed to establish an evaluation index system for the general layout of nuclear power plants. By comprehensive analysis of the main influencing factors of the project such as the self-human condition, three overall layout schemes are quantitatively evaluated from each dimension, and the optimal scheme is obtained.
The Research on Common Cause Failure Analysis and Countermeasures for Nuclear Safety Ventilation and Air Conditioning System
Authors: Peng Yongsen , Liu Zhengjie 
Published: November 2022

Abstract
By studying relevant regulations, standards and good practices, this paper discusses the typical types of common cause failures (CCFs) and general countermeasures in nuclear power plants, and summarizes the methodology, process and defenses for CCFs analysis of safety ventilation and air-conditioning (VAC) system in nuclear power plants. The typical CCFs are divided into four categories, namely functional dependency, spatial dependencies, inherent dependencies and human-related dependencies, and the corresponding defences are functional isolation, physical isolation, and diversity respectively. Comprehensive CCFs risk analysis of VAC system involves plant fault study, equipment FMECA analysis, PSA analysis, hazards analysis and other fields. For the identified CCF points, The optimal solution can be determined by combining factors which include technology maturity, feasibility, engineering cost, etc., and methods such as equipment diversity, physical isolation and functional isolation can be used to reduce or eliminate CCF points, to achieve as low as reasonably practicable (ALARP) risk of CCFs. The CCFs analysis methodology, process and defences summarized in this paper can provide reference for the CCFs of other nuclear safety supporting and auxiliary systems in nuclear power plants.
Analysis on the Determination of Coastal Nuclear Power Plant Platform Level
Authors: Huang Chong Guang , Wu Yan 
Published: November 2022

Abstract
In the project feasibility study stage, the determination of nuclear power plant factory dikes can effectively promote the rapid progress of the project, which can accurately calculate the quantity of construction conditions, the concrete position of the artificial slope, and other topics about factory ping elevation design input, effectively promote the progress of each project more forcefully demonstrates the rationality of the nuclear power plant engineering plan. It is very important to determine the plant floor elevation efficiently and reasonably. Based on this significance, a comparison and selection model for determining the plant floor elevation of nuclear power plant is established.
In the project feasibility study stage, the reasonably determined floor elevation will be used as the design input of other related demonstration topics, and there is a process of mutual iterative demonstration. The floor elevation of nuclear power plant needs to be determined based on various conditions of the site, so as to achieve the purpose of reducing the engineering cost and operation cost of nuclear power plant. This paper introduces the process of determining the floor elevation of a coastal nuclear power plant, and determines the floor elevation of a certain coastal nuclear power plant through comparison and selection analysis model. This analysis model is universal and important, and can provide the demonstration method and basis for determining the floor elevation of other coastal nuclear power plant sites. The analysis model mainly involves such factors as the amount of earthwork, the operating cost of circulating cooling water, and the slope of large transportation road.
Case Study on Significance Determination Process for NPPs Abnormal Events at Power
Authors: Chen Shijun , Zou Xiang , Chen Lihui , Luo Wenbo 
Published: November 2022

Abstract
With the research and application of Significance Determination Process (SDP) in China, it is urgent to carry out the detailed analysis of the significance of abnormal events at power, especially the NRC reactor regulation concept is being accepted to regulate the nuclear power plant in china. Based on Significance Determination Process (SDP) and the actual case analysis, this paper analyzes the significance of license operation event in a nuclear power plant with the current developed PSA model. The analysis process includes the initial significance determination, detailed analysis and external events/LERF analysis, finally determine the significance of license operation event, and provide a reference for the subsequent analysis of abnormal events.
Critical Heat Flux Experiments for IVR-ERVC Strategy Under the Pool Boiling Condition
Authors: Wang Gang , Kuang Bo , He Yihai , Liu Pengfei 
Published: November 2022

Abstract
Heat transfer limit is one of the main concerns of IVR-ERVC strategy. When the liquid level in the system is so low that natural circulation cannot be formed, the coolant near the outer surface of the reactor pressure vessel lower head is in the pool boiling state. In this research a one-dimensional full-height experimental facility was established to research the heat transfer limit (CHF) of ERVC under the pool boiling condition with a one-dimensional heating block, which is used to simulate the lower head of reactor pressure vessel. The experiment was carried out at different liquid levels and the results are compared with those of natural circulation experiment at the same liquid level. Experimental results show that CHF increases with the increase of the inclination angle of heating block. Meanwhile, the increase of liquid level is beneficial to the improvement of CHF. In addition, it can be found that the influence of flow path size on CHF is complex, and CHF does not change monotonically with the increase of flow path size. On the other hand, compared with the results of natural circulation at the same liquid level, the CHF values under pool boiling are relatively low. These results are expected to improve the understanding of IVR-ERVC strategy.
Review of Flow Characteristics in Narrow Rectangular Channels Under Marine Conditions
Authors: Wang Enpei , Li Lei , Wei Mingjun , Wen Yongsheng , Zhou Bowen 
Published: November 2022

Abstract
With the extensive application of offshore power plants and attaching importance of nuclear reactor system in marine industry, the narrow rectangular channels used for the compact heat exchanger in marine nuclear plant have drawn growing attention in recent years. Compared to some conventional pipes, the narrow rectangular channels boost the heat removal capacity efficiently and present the characteristics of compact layout, which have extensive application in compact heat exchangers. A significant number of experiments have been conducted on this topic. In this paper, a comprehensive historical review for narrow rectangular thermal-hydraulic characteristics under marine conditions is proposed, including an expansive number of experimental and numerical studies in open literatures for recent decade. The pressure drop and heat transfer characteristics are chiefly clarified in this paper. Additionally, the closely related studies of authors published in the last years are summarized in detailed. The followings are the significant aspects of this paper for dominant contents: (1) on the basis of a compilation of numerous distinguishing criterions, the definitions of narrow rectangular channels and marine motions are presented. (2) the flow characteristics of pressure drop and heat transfer characteristics in narrow rectangular channels under marine conditions are reviewed and summarized, basing on the authors’ published research achievements.
Integration and Future Development of Technical Standards for PWR Nuclear Pressure Equipment
Authors: Wu Feifei , Dong Zhenbang , Huang Jun , Liu Shangyuan 
Published: November 2022

Abstract
The development of nuclear power safety and the global procurement of nuclear power equipment have put forward an urgent need for the integration of nuclear power standards. In this paper, the development history and technical status of the localization of pressure equipment in China’s PWR nuclear power plants is introduced. This paper expounds the technology innovation and engineering experience of China’s independently-developed third-generation PWR nuclear power based on “HPR 1000” and “CAP 1400”, and the main goals, principles and basic ideas of the integration of technical standards for nuclear pressure equipment by sorting out and analyzing the list of topics in the fields of materials, design, manufacturing, welding, non-destructive testing and other professional fields involved in the integration of China’s nuclear pressure equipment standards, comparing with the technical topics involved in the current work of the SDO Integration Committee, and further discussing the differences and connections. This paper discusses the future development trend of nuclear pressure equipment in pressurized water reactor nuclear power plants and the demand for standardization of key technologies, and proposes the main problems that should be paid attention to and solved in the next step in the integration of nuclear pressure equipment standards.
Study on Fatigue Properties of Hydrided CZ2 Alloy
Authors: Huang Jieming , Gao Changyuan , Chen Liutao , Wang Xu , Tan Jun 
Published: November 2022

Abstract
CZ2 alloy is a new Zr-Nb zirconium alloy independently developed by China Nuclear Power Research and Technology Institute (CNPRI), which has independent intellectual property rights. Zirconium will react with coolant to absorb hydrogen in reactor service, which will degrade the performance of zirconium alloy and lead to failure. Therefore, in many studies of zirconium alloy performance, hydriding is often used to simulate burnup. In this paper, CZ2 alloy with hydrogen content of about 400 ppm was obtained by gaseous hydrogen charging method. The sample was heated to 500°C with the furnace in hydrogen environment and kept for 1.5 h. Then the fatigue tests of hydrided/un-hydrided CZ2 alloy and un-hydrided Zr-4 alloy under asymmetric axial pull-pull loading at 343°C were carried out to study the effect of hydride on the fatigue properties of CZ2 alloy. The results show that with the decrease of stress level, the difference between the fatigue life of CZ2 alloy un-hydrided and that of hydrided becomes larger, and the fatigue life of CZ2 alloy un-hydrided is higher than that of hydrided. The fatigue properties of un-hydrided CZ2 alloys are better than Zr-4 alloys. The two-parameter model and three-parameter model are used to fit the fatigue data of CZ2 alloy. It is found that the three-parameter model can better describe the fatigue life of CZ2 alloy.
Research on the Public Radiation Cognition Based on Social Psychology
Authors: Li Guanghui , Gong Yu , Bai Gang , Qu Yunhuan , Li Xiaoding 
Published: November 2022

Abstract
Radiation has the characteristics of being silent, colorless, odorless, invisible, and intangible. The public is deeply fearful and worried about radiation. Radiation has become “to mention the wolf’s name is to see the name”. Carrying out radiation science popularization is an effective way to crack the public’s fear. Research on public cognition of radiation is the basis of science popularization. Social psychology is a branch of psychology that studies public behavior dominated by psychology. Using social psychology can effectively analyze the psychology of public radiation cognition and improve the pertinence and effectiveness of science popularization. This paper investigates the current situation of public radiation cognition, and analyzes the existing problems of radiation cognition from the perspective of social psychology. It mainly includes that the public’s first impression of radiation is wrong, the public’s cognition of radiation is one-sided, ignoring the public’s needs, underestimating the public’s sense of resistance to radiation, underestimating the influence of wrong information, etc. In view of the above problems, this paper puts forward the methods to strengthen the pertinence and effectiveness of radiation cognition from the aspects of enhancing the effect of persuasion, paying attention to public needs, early public participation in radiation activities and reducing the impact of wrong information, so as to provide reference for radiation science popularization.
Study on the High Temperature Melting Treatment of Nuclear Waste Glass Fiber
Authors: Liu ChunYu , Wen YanLi 
Published: November 2022

Abstract
In abroad, the treatment method adopted for the wastes above is mechanical compression and storage [1–2]. In domestic, the wastes management methods could be summarized as follows according to the surface dose of the waste substance: (1) Waste filter cores with a surface dose rate greater than 2 mSv/h are fixed in a 400L steel drum with cement in the cement solidified line of nuclear auxiliary plant (NX), and one waste filter core is installed in a 400L drum [3–5]. After reconditioning, the waste volume is 0.4m3; By storing decay, the surface dose rate of some of the high dose rate level waste filter cores decreased to below 2 mSv/h, and then they were dried, super-compacted and cemented, which greatly reduced the amount of waste generated. (2) The waste filter elements with the surface dose rate below 2 mSv/h are packed into 200L steel drums and sent to the waste treatment auxiliary workshop (QS) for drying, super compaction and cement fixation. About 3∼4 waste filter elements are packed into a 400L drum. After preparation, the average waste volume of each filter core is 0.1m3. However, the treatment time of these methods is very long, the volume of waste after treatment is still relatively large, and the storage and isolation time is long. To sum up, this paper innovatively adopted high-temperature melting method to conduct glass solidification treatment on the simulated contaminated glass fiber, and by this method it forms a stable glass body with stable chemical properties. Moreover, it reduces the waste volume and directly forms a solidified body which is more convenient for treatment and disposal.
Analysis and Research of Code Revision for Design of Nuclear Safety Related Concrete Structures of Pressurized Water Reactor Nuclear Power Plants in China
Authors: Huang Jun , Zhang Weiguo , Liang Xueyuan , Wu Feifei , Xu Jing , Zhang Shaolin , Yang Wen , Tang Zhiyao 
Published: November 2022

Abstract
NB/T 20012-2010 design requirements for concrete structures related to nuclear safety of PWR nuclear power plants is a standard issued by the National Energy Administration in 2010. It is the top-level standard and important standard for concrete structure design. Due to technological progress and changes in relevant reference standards, some of its contents can not meet the requirements at this stage, On the basis of referring to the latest relevant standards and specifications at home and abroad, the engineering experience in the structural design of nuclear safety related plants of nuclear power plants built and under construction in China was summarized, and the chief editor revised it. This paper mainly analyzes the revised contents, mainly including revising the minimum strength grade of concrete was changed to C30, HRB500, HPB300 steel bars were added, and HPB235, HRB335 steel bars were canceled; improving the safety of the axial compression members that configure the ordinary stirrups, modifying the formula for calculating the bearing capacity of the transverse shear (that is, out-of-plane shear), modifying the formula for calculating the punching bearing capacity (considering the effects of membrane force) ; defining the load effect combination used in the calculation of crack width, and the maximum crack width is calculated according to the relevant specifications of GB 50010; stipulating that short-term stiffness is taken in deflection calculation under normal operation and severe environment, normal operation and extreme environment and abnormal operation conditions; adding contents of structural analysis, basic provisions of mechanism components and beyond the scope of design basis;The Appendix B (informative) “Performance Index of Reinforcement and Concrete at High Temperature” was added;.
Applying Machine Learning to Improve BOA Analysis of CIPS/CILC
Authors: Liao Jun , Mace Alexander , Wang Guoqiang , Fisher Jesse S. 
Published: November 2022

Abstract
CRUD depositions on fuel cladding are generated as a result of coolant chemistry and corrosion of primary system components during the normal operation of a nuclear reactor. These depositions can affect both reactor power peaking and fuel element structural integrity through CRUD Induced Power Shifts (CIPS) and CRUD Induced Local Corrosion (CILC) phenomena, respectively. The CIPS/CILC phenomena are typically assessed using the Boron-induced Offset Anomaly (BOA) computer code, which predicts the characterization of the distribution of CRUD within a reactor and the risk for occurrence of CIPS/CILC within a given cycle.
Machine learning (ML) algorithms have often been applied towards understanding large data sets and acquiring useable information in many industries, including nuclear. In the nuclear engineering field, ML is increasingly being used as a method to improve nuclear power plant efficiency and reduce engineering costs. The application of machine learning to enhance the prediction of CIPS/CILC increases the possibility of reducing the cost associated with CRUD mitigation. In Westinghouse, exploratory analysis was performed to identify ways in which ML could be utilized to improve CIPS risk assessments. There were two paths identified, the first one was ML applications for CIPS with the standalone BOA code, and the second path was ML applications for CIPS calibration with coupled ANC/FUELUDTYDRV/VIPRE-W/BOA code. The exploratory study demonstrated that ML was capable of accurately predicting the boron mass in the cores of selected nuclear power plants as surrogate models of the BOA code.