bluuebunny/crossref_metadata_2025_split huggingface dataset:
Research and Application of Remote Intelligent Operation and Maintenance System for Digital Instrument and Control Equipment in Nuclear Power Plant
Authors: Guilian Shi , Xiaojin Huang , Xiaopeng Zhao , Lei Wang 
Published: November 2022

Abstract
In view of the problems of poor maintenance timeliness of nuclear power digital instrument and control system (I&C system) and the inability to evaluate the health condition of the I&C system in real time, a remote intelligent operation and maintenance strategy based on virtual DCS and Prognostic and Health Management (PHM) technology framework was proposed. The key technical information of virtual DCS and PHM is introduced. Combining with the technical characteristics of digital I&C system of nuclear power plant (NPP), the remote intelligent operation and maintenance framework combining virtual and actual object is constructed. With respect to the actual I&C system, virtual DCS was established to realize condition based maintenance for its flexibility and predictive characteristics. At the same time, the actual system data can be collected and stored through the I&C system data center, which will be the data basis for the remote monitoring of the site. Then, five functions of remote monitoring, fault diagnosis, condition prediction, intelligent maintenance and resource allocation and management are described. The key technologies and data resources needed to realize the scheme are analyzed. Finally, according to the strategy, a prototype with remote intelligent operation and maintenance function is built on the basis of a nuclear power platform, which realizes the strategy and provides feasibility guarantee for remote intelligent maintenance of I&C equipment.
Validation of NUIT for Decay Heat Predictions of PWR Spent Fuel Assemblies
Authors: Li Jian , Li Ruihan , She Ding , Liang Jin'gang , Shi Lei 
Published: November 2022

Abstract
Decay heat prediction is essential for nuclear reactor safety, waste management and spent fuel disposal. The NUIT code is dedicated for nuclide inventory calculations, which was developed by Institute of Nuclear and New Energy Technology (INET), Tsinghua University. NUIT is able to compute isotope density, radioactivity, decay heat, photon spectrum and other source term quantities. The problem-specific neutron spectrum and tabular effective one-group cross-section library produced by neutron transport codes were utilized by NUIT. Therefore, the NUIT code could be applied to many types of fission reactors, such as HTGR, PWR, BWR, etc. In this paper, validation of the NUIT’s capabilities in predicting decay heat for spent fuel assemblies was conducted with measured decay heat data from PWRs. The test cases for validation include measurements of 52 fuel assemblies (14 × 14, 15 × 15 and 17 × 17 fuel designs) that were irradiated in 5 PWR plants, with burnup values ranging from 19.7 to 51 GWd/tU. The cooling periods for the measured samples vary from 2.4 to 26.7 years, which are times of interest to spent fuel transportation and storage applications. The multi-group neutron spectrum provided by Monte Carlo code OpenMC was utilized to collapse into one-group cross sections. The decay heats for different fuel assemblies were obtained and compared with measurement data. The validation results show satisfactory agreement between NUIT results and reported measurement data. This work helps to obtain uncertainty margins in decay heat predictions with NUIT code.
Interpolation Influence on the Fast Fourier Transform Based Calculation of Three-Dimensional Dose Rate Field
Authors: Dong Xinwen , Fang Sheng , Zhuang Shuhan 
Published: November 2022

Abstract
The calculation of three-dimensional dose rate fields plays a key role in radiation dose rate estimation and the service for the nuclear emergency. The recent fast calculation method based on the Fast Fourier Transform (FFT) method can greatly speed up the calculation without losing accuracy, which is promising for operational usage in nuclear emergency response systems. But it can only be used for a uniform grid. Unfortunately, most atmospheric dispersion models use a non-uniform grid, which prevents the direct application of FFT-based calculation. Therefore, interpolation is required beforehand to use the Fourier transform, which may introduce errors and affect computing efficiency. In this paper, an atmospheric dispersion modeling case of a typical nuclear power plant (NPP) is used to investigate the efficiency of different interpolation methods, which are based on a non-uniform grid. These methods are linear interpolation and nearest-neighbor interpolation. The sensitive analysis of grid resolution is investigated in the slices of x, y, and z at typical positions, which confirms the smooth-out and speed-up effects in rough grids. A grid size over 10 m at any slice commonly causes losses of change details of dose rate fields. Given the same resolution of 50 m × 50 m × 50 m, the nearest neighbor performs a 717 times calculation faster than the linear method, which preserves more change details of dose rate fields as well. For complex calculation tasks, e.,g., non-uniform NPP buildings, the nearest neighbor interpolation method is recommended with a resolution of 10 m × 10 m × 10 m to make a good balance between accuracy and speed.
The Impact of Corporate Social Responsibility on Public Acceptance of Nuclear Power: A Mediation Model
Authors: Xu Zhong , Ma Jianhong , Li Xuning , Yang Ting , Gui Xinyu 
Published: November 2022

Abstract
Nuclear power, as a clean energy source, has been facing a blow after the Fukushima nuclear accident. Low public acceptance has severely restricted the nuclear power industry’s development. China government attaches great importance to the public acceptance of nuclear power and takes it as an important indicator for the approval of nuclear power projects. In the research of consumer domain, corporate social responsibility (CSR) is positively correlated with consumer support for enterprises, and the public is the ultimate consumer of nuclear power products. On this basis, hypotheses are proposed in this study: the public’s perception of the social responsibility of nuclear power enterprises can improve the public acceptance of nuclear power. Taking 211 residents around a nuclear power plant as the research object, this paper explores the impact of CSR perception on public acceptance of nuclear power and its path through questionnaire survey. The data analysis results of valid questionnaires showed that: (1) The percived CSR of nuclear power companies in China is divided into two dimensions, legal responsibility, economic&other responsibility, (2) The perceived legal responsibility has a positive effect on the public acceptance of nuclear power, and this effect is realized through the partial mediating effect of knowledge, dread risk perception and benefit perception.
Research on the Influence of Degassing on the Release of Fission Gases in PWRs
Authors: Ye Yaoxin , Fu Pengtao , Li Zhijun 
Published: November 2022

Abstract
Once fuel failures occur in a reactor core, the fission gas will release from the UO2 pellet to the fuel-to-sheath gap and eventually enter the primary coolant through the cladding wall breach. The crew of nuclear power plant regularly sample the primary coolant and characterize the fuel integrity based on the fission gases release model. If the activity of the primary coolant is too high and exceeds a certain limit, the degassing operation will be performed by the crew, causing a significant variation of primary coolant activity. As a result, the influence of degassing on the fission gases release should be taken into consideration in such situation. In this work, a modified model of fission gases release considering the degassing effect was established based on the fission product release mechanism and the authentic degassing processes of a nuclear power plant. The modified fission gases release model with degassing consideration was then validated by the measured primary coolant activity in a nuclear power plant. Finally, the model was applied on characterizing the fuel integrity of a fuel cycle with defective rods and frequent strong degassing operation. The dependence of release-to-birth ratio on decay constant of fission gases was determined using the modified model and the defect size was predicted. The predicted defect size was in good agreement with the sipping test result.
Active Interrogation of Highly Enriched Uranium in the Suitcase by Using NG-9 Neutron Generator
Authors: Shi Guang-Yu , Jing Shi-Wei 
Published: November 2022

Abstract
Preventing the proliferation of highly enriched uranium (HEU) is an important issue all over the world. The active interrogation of HEU hided in the suitcase based on neutron technique is studied by using MCNP5 code. The neutron source of the detection system is based on the NG-9 neutron generator developed by Northeast Normal University. A set of HEU detection devices has been established. Lead is a kind of neutron-multiplier material with high density, so the lead block is chosen as the substitute for HEU. The size of the aluminum alloy suitcase is 58 cm × 42 cm × 25 cm. The NG-9 D-T neutron generator is placed on the suitcase. A lead block measuring 26.5 cm × 12 cm × 1 cm is placed in the center of the suitcase. The lead block has a density of 11.35 g/cm and a mass of approximately 3.6 kg. Daily clothes are placed inside the suitcase as a distraction. A cylindrical BGO detector with a diameter of 7.2 cm, a height of 29.7 cm is placed close to the suitcase to record gamma rays. A cylindrical lead shield with a thickness of 5 cm is placed outside the BGO detector. Paraffin wax is placed around the whole detection device to protect the neutron radiation and avoid the interference of other substances on the detection results. The purpose of this experiment is to verify the agreement of the MCNP5 simulation results and the experimental results. In this paper, threshold energy neutron analysis (TENA), fast neutron method, and thermal neutron method are used to detect HEU. The simulation results show that the presence of HEU in the suitcase can be determined by neutron flux which is higher than those in the absence of HEU.
Study on Vertical Transfer of Radionuclide Co-60 in Broun Soil and Aeolian Sandy Soil Around Nuclear Power Plant
Authors: Zhang Qiong , Yuanlv Ye , Bo Wang , Lu Chen 
Published: November 2022

Abstract
The coastal soil surrounding to the Nuclear Power Plants were mainly classified to brown soil and aeolian sandy soil in China. In previous research, focused on the elements of Cs-137 and Sr-85 vertical transfer in these soils, but the Co-60, which was the most important factor in the brown soil and aeolian sandy soil were still urgent research. In this study, experiments were conducted to determine the vertical transfer of Co-60 in brown soil and aeolian sandy soil. The soil around Shidaowan Nuclear Power Plant were studied by using undisturbed soil column methods and isotope tracer technology, and identify the factors influencing their migration depths in soil. The results show that with the increase of leaching test, the migration of Co-60 in aeolian sandy soil was greater than that in brown soil, but the difference gradually decreases with the progress of spray test; the content of Co-60 in leaching water was very few and cannot be detected due to the strong adsorption by soil; after three years of undisturbed soil column experiment, 72.36 ∼ 85.26% of Co-60 mainly was retarded in the range of 0∼5cm in both of the soil surface. It was proved that Co-60 was difficult to migrate to groundwater in a short period. The distribution of specific activity of Co-60 in soil was an individual exponential declining with depth of oil. The results showed that the radionuclide Co-60 had little impact on groundwater in the years after the nuclear accident, but it should pay attention to the change of radioactivity concentration in surface soil and then select appropriate soil remediation methods. The study will provide a scientific basis for post-accident environmental impact assessment and post-accident soil remediation in China.
A Novel Transmission Reconstruction Algorithm for Radioactive Drum Characterization
Authors: Yang Hui , Zhou Hao , Dong Bing , Zhou Wentao , Gu Weiguo , Zhang Xinyu , Lei Qingxin , Shan Chenyu , Wang Dezhong 
Published: November 2022

Abstract
The accuracy of tomographic gamma scanning transmission reconstruction is a critical factor in reconstructing the activity of a radioactive drum. Traditional reconstruction algorithms produce severe grid artifacts and a high level of noise, thereby increasing the reconstruction error for both the density map and the activity. This paper proposes a novel algorithm for transmission reconstruction by combining maximum-likelihood expectation maximization and a convolutional neural network (CNN). Our experimental results indicate that the proposed reconstruction algorithm is capable of significantly reducing measurement errors, increasing spatial resolution while also eliminating grid artifacts, and being sufficiently robust when dealing with a noisy input image. The mean squared error of the output image decreased by 52.70% compared with the conventional reconstruction method, and the peak signal-to-noise ratio and structural similarity index improved by 21.89% and 17.33%, respectively. The spatial resolution was improved by 28 times, which demonstrates that CNN is a potentially useful new method for radioactive waste drum transmission image reconstruction.
Research on Risk Management of Nuclear Engineering - Taking T Project Reverse Power Supply As an Example
Authors: Wang Baoquan 
Published: November 2022

Abstract
Under the vision of sustainable development and a community with a shared future for mankind, the era of “zero carbon” is coming to us. Nuclear energy has long been seen as one of the powerful means to achieve a “carbon neutral” future. Nuclear power projects are characterized by high technical requirements, long construction period, complex system composition, numerous stakeholders and heavy social responsibilities. Any link is crucial. Therefore, effective risk management of nuclear power projects is of great significance.
Based on the actual situation of nuclear POWER T project, the following studies are carried out:
(1) Through the research of domestic and foreign literature, Delphi method anonymous voting and expert consultation meeting were adopted to merge, filter and supplement the risk factors, and 24 risk factors in seven categories, including design, installation and commissioning, construction, contract, external, procurement and administrative licensing, were identified in the process of power transmission of NUCLEAR POWER T project.
(2) AHP fuzzy comprehensive evaluation method is used to select the hierarchical structure model of power supply risk factors, construct a judgment matrix, establish a risk evaluation table, calculate the risk weight of each level and rank it, compare the value interval of risk evaluation set, and get the power supply risk level of NUCLEAR power T project.
(3) In the outlook, the author considers the use of bow-tie model to prevent and control risk events in advance in the future, so as to formulate measures. Finally, the integrated risk control platform of reporting, early-warning, disposal and monitoring is proposed to form a closed-loop management of risk identification, reporting, disposal and rectification.
Application of the Refinement Spacer Grid Model in Subchannel Analysis Code
Authors: Dong Si Ying , Yang Ping , Gan Fujun , Liu Yuanrong 
Published: November 2022

Abstract
As one of the important components in PWR cores, spacer grids can maintain the position of the rods during operation as well as promote fluid motion. The overall physical phenomena and local effects introduced by spacer grids can be accounted for the crossflow mixing model. Based on the in-house thermal-hydraulics (T/H) subchannel code SAPRE, the four improvements are developed:1) The distributed resistance method (DRM) is applied to develop the crossflow mixing model in order to reflect mixing vanes grid’s contribution to crossflow mixing in rod bundles. 2) By taking more detailed grid parameters into account, the loss coefficient of grid is calculated based upon form and friction losses. 3) The Gunter-Shaw frictional relationship is established to simulate the transverse resistance caused by the spacer mixing vanes. 4)Improvement of lateral momentum equation. Due to the presence of the mixing vanes, the crossflow is increased, and the domination of axial flow in the lateral momentum equation is not appropriate. The momentum changes caused by the crossflow on the flow term should be considered. All these four models are implemented in the subchannel code SAPRE. The improved SAPRE code is validated through Hassan 5 × 5 Particle Image Velocimetry experiment and PSBT benchmark. Numerical results demonstrate that SAPRE has large advantages in accurate prediction of local T/H parameters in rod bundles, which is of significance to safety and economy of PWR.
Microstructural Evolution of SiC-Coated C/C Composites Under Argon Ion Irradiation
Authors: Xie Xiangmin , Yan Long , Cheng Guodong , Tang Xian 
Published: November 2022

Abstract
SiC coatings have been used to improve the oxidation resistance and stability of C/C composites in high-temperature reactors. However, the irradiation-induced surface structural transformations of SiC-coated C/C composites have been rarely studied. Herein, chemical vapor reaction (CVR) SiC-coated C/C composites were irradiated with 300 keV argon ions at room temperature with irradiation doses ranging from 5 × 1015–1 × 1017 ions·cm−2. The damage patterns of the pristine C/C composites and SiC-coated C/C composites were observed using scanning electron microscopy, and the shape and size evolutions of the CVR-SiC particles were investigated as a function of the irradiation dose. The results revealed that the pristine C/C composites were severely damaged after ion irradiation, and a large number of defects and pores formed on the surface. In contrast, the ion-irradiated SiC-coated C/C composites showed an undamaged surface. As the irradiation dose increased from 0 to 1 × 1017 ions·cm−2, the CVR-SiC particles were transformed from irregular to spherical shapes, and the average size of the SiC particles was reduced from 22 to 5 μm. The size reduction and spheroidization of the SiC particles under irradiation were attributed to the amorphous transformation of SiC. This study can provide deeper insight into the irradiation behavior of SiC-coated C/C composites in high-temperature reactors.
On the Creation of Nuclear Law as a New Inter-Discipline in Law
Authors: Liu Jiu 
Published: November 2022

Abstract
Civil nuclear industry has developed rapidly and made remarkable achievements all over the world since 1940s. The nuclear legal system plays an important role in promoting and guaranteeing the development of civil nuclear industry and utilization of nuclear energy. The implementation and improvement of nuclear legal system form the corresponding discipline of nuclear law. The construction of nuclear law as a new inter-discipline of both nuclear science and technology and law needs a timely follow-up. Therefore, this paper starts from the analysis of the nuclear law as a branch law and studies the nuclear law as a new inter-discipline in law. Nuclear laws is still underway and needs to be improved currently. In order to create the new inter-discipline of nuclear law, it is necessary to make clear the definition, object and characteristics of the discipline, and to design well in terms of teachers and curriculum so as to make the discipline of nuclear law more mature and realize its value as a new inter-discipline in law.
Research on the Design Method of Nuclear Heat Exchanger Tube Based on Multi-Objective Optimization
Authors: Cao G. M. , Dai L. , Zhao Z. X. , He T. 
Published: November 2022

Abstract
In addition to the traditional performance indexes such as heat exchange efficiency and flow resistance, the design of nuclear heat exchange tube needs to take into account the flow noise level index of heat exchange tube at the same time. Therefore, the design of heat exchange tube needs to consider multiple properties such as heat exchange efficiency, flow resistance and flow noise at the same time, which greatly increases the complexity of nuclear heat exchange tube design. In this paper, a design method of low-noise nuclear heat exchange tube is proposed. The design parameters of nuclear heat exchange tube are automatically adjusted based on the coupling of multiple simulation software. The heat exchange efficiency, flow resistance and flow noise of nuclear heat exchange tube are characterized by normalizing the comprehensive evaluation factor of flow and heat transfer and insertion loss, and the optimal design parameters of nuclear heat exchange tube are found through orthogonal test.
Criticality Safety Analysis of Fuel Storage Under Fuel Assembly Dropping Accident
Authors: Dai Mingliang , Fu Pengtao , Zhao Jun 
Published: November 2022

Abstract
During fuel storage, lifting fuel assembly is an essential process. The fuel assembly dropping accident may lead to consequences, such as fuel assembly lying on the storage rack, fuel assembly deformation leading to out of control of fuel rod pitch distance, fuel rod breakage and other accidents, and then affect the critical safety of fuel storage system. Therefore, it is necessary to analyze the critical safety of fuel assembly storage under fuel assembly dropping accident.
This study analyzes all kinds of accident conditions that may occur under fuel assembly dropping accident, and uses Monte Carlo particle transport code JMCT to simulate the neutron effective multiplication factor (keff) at typical conditions. The results show that the calculated keff of the fuel storage system will not exceed the criteria when the fuel assembly drops on the fuel storage rack without deformation. When fuel assembly is deformed due to dropping from a height, change of fuel rod pitch distance in fuel assembly will affect the moderator-uranium ratio of the fuel assembly storage system, which may lead to an increase in keff in the system. However, the fuel storage system can also will not exceed the criteria when 2000 ppm soluble boron acid is taken credit in the spent fuel pool. Based on the above analysis, the fuel storage system can ensure the subcritical state at different condition of fuel assembly dropping.