bluuebunny/crossref_metadata_2025_split huggingface dataset:
Proposal of Detailed Procedures of Determining Rational In-Service Inspection Requirements Based on System Based Code Concept
Authors: Yada Hiroki , Takaya Shigeru , Morohoshi Kyoichi , Yokoi Shinobu 
Published: November 2022

Abstract
In order to develop rationalized maintenance plans of nuclear power plants, it is necessary to consider characteristics of each plant. For sodium-cooled fast reactor (SFR) plants, there are constraints on inspections due to the specialty that sodium equipment needs to be handled, which is one of the important points when considering rationalization of maintenance. Therefore, we previously proposed a maintenance optimization scheme based on the System Based Code (SBC) concept. The concept of the SBC intends to give a theoretical procedure to optimize the reliability of the system by administrating every related engineering requirement throughout the life of the system from design to decommissioning. “ASME Boiler and Pressure Vessel Code, Code Case, N-875 Alternative Inservice Inspection Requirements for Liquid-Metal Reactor Passive Components” was already developed based on the SBC concept. The basic procedure of the code case has been also implemented in “ASME Boiler and Pressure Vessel Code, Section XI, Division 2” which provides requirements for Reliability and Integrity Management (RIM) programs for nuclear power plants including advanced reactors. One of proposed scheme goals is to develop detailed procedures of preparing a maintenance plan based on the process in the code case. In this study, the procedures to determine inspections for potential degradation and additional inspections in terms of defense-in-depth have been further clarified. The modified maintenance optimization scheme consists of the following processes. At first, potential active degradation mechanisms are identified. If there are any degradation mechanisms that are not addressed in the design codes, structural reliability evaluations are conducted to confirm whether the component of interest has sufficient reliability for the degradation mechanisms of concern. In the case the evaluated reliability does not meet a predetermined reliability target, inspections for the potential degradation are required. Next, the detectability of hypothetical defects before they grow to an unacceptable size in terms of plant safety is assessed. A method which has sufficient detectability, either in direct way such as volumetric examinations or in indirect way such as monitoring, is required as an additional inspection. If any methods are not applicable, reliability evaluation with a penalty is required. The penalty is determined conservative hypothesis relating safety function of the component of interest. If the evaluated reliability meets to the reliability target, additional inspections can be exempted. In this paper, the modified maintenance optimization scheme is also illustrated by a quantitative trial evaluation of the core support structure of the next SFR under development in Japan.
Research on Improving the Science Popularization System of Nuclear-Related Projects in China
Authors: Zhang Zhenhua , Zhu Li 
Published: November 2022

Abstract
Nuclear power itself has the characteristics of safety, cleanliness and high efficiency, but due to the professional particularity of nuclear energy and the people’s lack of knowledge about nuclear energy, it leads to the “social amplification of risks” in nuclear-related projects. Some basic data in the discipline of nuclear safety and radiation protection are analyzed, and the effects of the nuclear power chain on health, environment and climate were compared with those of the coal power chain and then the advantages of the nuclear energy industry are given. The article also compares and analyzes the nuclear energy risks and other risks, and analyzes the reasons for the public misunderstanding and fear of nuclear and radiation accidents (events). Now there are many problems existing in the nuclear science popularization work, such as the trend that the risk of nuclear-related projects is easy to be amplified by the society hinders the process of the nuclear energy industry to some extent. Also there are many deficiencies of support and popularization, and many problems in the development of high-quality popular science teaching materials and models and the cultivation and sharing of high-quality resources. Based on the above analysis, it is proposed that nuclear power science popularization should make precise efforts, distinguish groups, take measures according to people, classified implementation and precise communication. Specific suggestions and measures have the following aspects: Multilevel nuclear science education should be carried out and it includes nuclear science for nuclear experts and non-nuclear experts (including educators), nuclear science for government staff and public opinion media, and nuclear science for the public; At the same time, it is necessary to improve the form of nuclear science education, and we will actively and extensively carry out popular science activities that are practical, close to the people, distinctive and effective and establish a long-term nuclear science education mechanism; Drawing on foreign experience to accelerate the formulation of China’s atomic energy law, and clarify the status of public participation in the law. So as to improve the legal and regulatory system, improve the public participation system, and strengthen the capacity building of the information disclosure platform.
Experimental Study on Dynamic Tensile Strength of Graphite
Authors: Tian Dongqing , Shi Li , Sun Libin , Shen Ke , Xu Kun 
Published: November 2022

Abstract
Dynamic tensile strength is an important parameter in the design of graphite components for High Temperature Gas-Cooled Reactors (HTGR) to evaluate the integrity of core support structures. The Digital Image Correlation and Split Hopkinson Pressure Bar (DIC-SHPB) test system was used to perform the disc compression tests according to ASTM D8289-20 to study the dynamic splitting tensile strengths of graphites with different grain sizes. The fracture process was captured by a high-speed camera and the tensile strain was computed by DIC. The results show that the SHPB test method is capable of performing the disc compression tests of graphite. The dynamic tensile stress-strain curve of graphite underwent four stages: compression stage, near-elastic stage, crack development stage and crack non-stable extension stage. When the strain rate is in the range of 330 s−1 to 630 s−1, the dynamic tensile strength of graphite increases with increasing strain rate. The dynamic increase factor (DIF) of fine-grained graphite ranged from 1.05 to 1.2, while that of coarse-grained graphite was about 1.2 to 1.6. There is a large dispersion in the dynamic tensile strength of graphite when the strain rate is increased. It was found that the tensile strength of graphite improved considerably with increasing strain rate, while the fracture strain decreased slightly.
Study on Analysis Method of LOCA Dynamic Response for Steam Generator Heat Transfer Tube
Authors: Lu Xifeng , Xiong Furui , Wang Xinjun , Ai Honglei , Li Bingjin , Shao Xuejiao , Yang Kang 
Published: November 2022

Abstract
The equivalent effect relationship formula of LOCA hydraulic load in the bend section of the heat transfer tube of the steam generator is deduced. The hydraulic load factors affecting the stress of the heat transfer tube are derived. The spectrum analysis of the corresponding LOCA hydraulic load is carried out, and the spectral curve of the hydraulic load and the spectral characteristics of the hydraulic load are obtained. Modal analysis of steam generator heat transfer tubes is done. The modal analysis of the steam generator heat transfer tubes is performed. The spectral characteristics of the hydraulic load and the natural frequency information of different heat transfer tubes are compared. According to the comparison results, eight heat transfer tubes with the first-order natural frequency and the frequency corresponding to the peak value of the hydraulic load in the frequency domain curve were selected as the analysis objects, and the LOCA dynamic response analysis is carried out. The time-domain analysis of the LOCA hydraulic load of the steam generator heat transfer tube was performed, and the relationship between the hydraulic load and the bending radius of the heat transfer tube was obtained. The transient load of the steam generator heat transfer tube with different bending radii is calculated. By establishing the finite element models of the first, 31, 43, 57, 72, 83, 87, and 112 rows of heat transfer tubes, the time history analysis of the heat transfer tubes under the LOCA load is performed, and The stress distribution of the heat transfer tube under the LOCA load is obtained. The stress of the heat transfer tube calculated by this method is the maximum LOCA response. This method can screen out sensitive heat transfer tubes, avoiding the process of analyzing each heat transfer tube. This method can effectively improve the efficiency of LOCA analysis of heat transfer tubes and is a practical method.
Influence of Low Coolant Level on Shell-Side Heat Transfer Performance in a Tubular Heat Exchanger
Authors: Liang Hui , Hu Zongwen , Cao Xiaxin 
Published: November 2022

Abstract
A passive residual heat removal heat exchanger is used in several nuclear reactors to remove reactor core decay heat after severe accidents. In the late stages of a severe accident, the coolant level in a water storage tank may decrease and heat transfer tubes cannot be fully submerged by coolant. To study the shell-side heat transfer performance at different low coolant levels, an experimental setup is built and experimental cases with four different coolant water levels are designed (i.e., full, 4/5, 2/3, and 1/2 coolant levels). During the experiments, saturated steam flows through the inner side of heat transfer tubes while purified water works as a coolant on the outside of tubes. According to the experimental phenomena, the shell-side heat transfer area can be divided into three regions, i.e., nucleate pool boiling region, liquid film evaporation region, and steam convection region. A new calculation method is proposed to calculate the shell-side heat transfer coefficients in different regions at low coolant levels. Experimental results show that the shell-side heat transfer performance is enhanced at 4/5 coolant level compared to the full coolant level case, similar to the full coolant level case at 2/3 coolant level, and lower at 1/2 coolant level. The heat transfer performance does not decrease proportionally to the coolant level reduction at low coolant levels because liquid film and small droplets impact the tube surface and phase-change heat transfer happens in the liquid film evaporation region.
Local-Scale Atmospheric Dispersion Modelling of Radionuclides Following the Fukushima Daiichi Nuclear Accident Using SWIFT-RIMPUFF
Authors: Zhuang Shuhan , Fang Sheng , Dong Xinwen 
Published: November 2022

Abstract
Air dispersion modeling is an important tool for emergency response following a nuclear accident, such as the Fukushima accident. Current researches mainly focus on global- and regional-scale modeling with wind field data derived from different methodological models and observations. However, the capability of the local-scale atmospheric dispersion model hasn’t been discussed in detail for the Fukushima accident. In this paper, the local-scale modeling of radionuclides following the Fukushima accident was investigated with the combination of the wind diagnosed model SWIFT and the radionuclide transport model RIMPUFF. The coarse input wind field of SWIFT is prepared with WRF using the European Centre for Medium-Range Weather Forecasts (ECMWF) Meteorological data. The SWIFT-diagnosed wind field was used to drive RIMPUFF for calculating the dispersion of radionuclides and gamma dose rates around the Fukushima Daiichi Nuclear Power Plant (FDNPP) site. The diagnostic wind fields were validated to the on-site meteorological observations, whereas the dispersion and gamma dose rates were validated the onsite gamma dose rate monitoring data and the observations of suspended particulate m? near the FDNPP. The plume patterns were also analyzed to help understand the transport behaviour of the radionuclides. The validation demonstrates that, with the 1km-resolved ECMWF meteorological data, SWIFT fairly reproduces the wind field. The speed is slightly overestimated, with a Normalized Mean Squared Error (NMSE) below 6. The wind direction is well simulated at some specific moments, which is critical for reproducing some peaks of the dose rates. However, RIMPUFF underestimates the dose rates around the FDNPP, partly because of the overestimated wind speed. The concentration simulations better agree with observations in the Naraha station than Futaba station, with the Fractional Bias (FB) under 0.3 and NMSE under 6 at the Naraha station.
Preparation Method of ORIGEN2 Library for High Temperature Gas-Cooled Reactors
Authors: Simanullang Irwan Liapto , Fukuhara Katsuki , Morita Keisuke , Fukaya Yuji , Ho Hai Quan , Nagasumi Satoru , Iigaki Kazuhiko , Ishitsuka Etsuo , Fujimoto Nozomu 
Published: November 2022

Abstract
The ORIGEN2 code has been used for fuel depletion calculations of many kinds of reactor fuels but there is no library for high temperature gas cooled reactors (HTGRs). A set of the ORIGEN2 library for the HTGR has been established to evaluate the fuel burnup characteristics. In this study, the ORIGEN2 library was prepared for the high temperature engineering test reactor (HTTR). The HTTR is the first Japanese prismatic type HTGR. The burnup dependent neutron spectrum is necessary for generating the ORIGEN2 library. A pin-cell burnup calculation was conducted to obtain the burnup dependent neutron spectrum in the fuel compact of HTTR. Then, the ORIGEN2 library was generated based on the neutron spectrum of the pin cell model. The calculation results that were calculated by the ORIGEN2 code was validated by comparison with a detailed calculation with use of the MVP-BURN code. This code-to-code method was used to validate the ORIGEN2 code calculation because of no assay data of HTTR spent fuels. One of the isotopes that evaluated was 239Pu. The calculation results showed that the amount of 239Pu calculated by ORIGEN2 code was higher about 35% than that of calculated by the MVP-BURN code. It showed that the ORIGEN2 library using the neutron spectrum of a pin-cell burnup model was not enough for evaluating burnup characteristics of the HTTR. Therefore, an improvement was performed to evaluate the ORIGEN2 library. In this study, the ORIGEN2 library was generated based on the neutron spectrum of a core burnup calculation. The calculation results showed that the ORIGEN2 code and the MVP-BURN code was in a good agreement. The maximum difference of 239Pu amount between the ORIGEN2 and MVP-BURN became 2.4%.
Research on Risk-Informed Regulation Approaches of MSPI and SDP
Authors: Zhang Guoxu , Chen Lu 
Published: November 2022

Abstract
By integrating operating experience, accident analysis, risk assessment and defense-in-depth philosophy into decision-making consideration, risk-informed regulation better focuses regulatory attention to an issue consistent with its risk significance, which facilitates the establishment of appropriate regulation without damaging nuclear safety at the same time. By referring to the inspection findings from nuclear power plant, this paper utilizes Mitigating System Performance Indicator (MSPI) and Significance Determination Process (SDP) for research. From perspectives of basic theory, risk analysis and regulatory application consideration, this paper analyzes the correlation and differences between MSPI and SDP. The work could provide reference for the application and promotion of Risk-informed Regulation Approaches.
Research on the Application of Cobalt-Based Alloy Laser Additive Manufacturing in Reactor Internals
Authors: Wang Qingtian , Cao Rui , Zhang Xianjun 
Published: November 2022

Abstract
In view of the quality problems of manual TIG coating of cobalt-based alloy, the Laser Additive Manufacturing process was studied, especially the optimization of process parameters. Then a series of experiments were carried out on the cobalt-based alloy Additive Manufacturing layer and compared with the traditional manual TIG coating layer. The results show that the microstructure, the hardness, the wear resistance and corrosion resistance are better than.
Corrosion Behavior of Zr-0.75Sn-0.35Fe-0.15Cr-xNb Alloys in Oxygen-Containing Super-Heated Steam at 400 °C
Authors: Yao Meiyi , Xu Shitong , Pei Wen , Hu Lijuan , Lin Xiaodong , Xie Yaoping , Peng Jianchao , Liang Xue , Zhou Bangxin 
Published: November 2022

Abstract
Due to a compact layout design of water-cooled small modular reactors (SMRs), no deaeration facilities is generally used, resulting in a high dissolved oxygen (DO) lever in the primary loop water. Therefore, the effect of DO on the corrosion behavior of zirconium alloys as fuel cladding materials is one of the issues that need attention in SMRs. In this study, the corrosion tests of Zr-0.75Sn-0.35Fe-0.15Cr-xNb (x = 0, 0.15, 0.3, 0.5, 1.0, wt%) alloys were carried out in super-heated steam with deaeration, 300 ppb DO and 1000 ppb DO at 400 °C/10.3 MPa using static and dynamic autoclaves. The microstructure and valence state of elements of the alloys and oxide films were characterized using TEM and XPS. Results show that DO accelerates the corrosion of Zr-0.75Sn-0.35Fe-0.15Cr-xNb alloys and the accelerated corrosion effect increases with the increase of Nb in the alloys and the DO content in the steam. Both the Nb in the alloys and the DO in the steam promote the oxidation of Fe to Fe2+ and Fe3+, and Cr to Cr3+, which will increase the number of defects to promote the diffusion of O2− vacancies, leading to accelerating the corrosion of the alloys. Compared with the deaerated environment, the transition layer at the oxide/metal interface formed in the DO environment is thinner, which provides less time for stress release during oxidation to promote the generation of cracks and accelerate the corrosion of zirconium alloys.
Numerical Simulation of Fuel Relocation at High Burnup With Discrete Element Method
Authors: Ma Zehua , Liang Ren , Lin Zhikang , Ouyang Yong , Lu Xianghui , Su Guanghui 
Published: November 2022

Abstract
In order to economically increase cycle lengths of pressurized water reactor, current regulatory burnup limit of 62 GWd/tU is expected to extend to ∼75 GWd/tU. According to the experimental observation in the Halden IFA-650 tests, high burnup fuel seems more susceptible to turn into fine fragments (pulvers) in loss-of-coolant accidents. Due to the cladding ballooning under temperature transients, the fragmented fuel can axially relocate into the ballooned region, which may lead to a local increase in cladding temperature and oxidation level, raising the licensing concerns related to core coolability as described in 10 CFR 50.46. Although fuel relocation has been extensively observed and well identified in various test programs, there is still data gap between experimental observation and numerical prediction. This paper proposes a three-dimensional simulation framework with discrete element method for fuel relocation, in which coarse fuel fragments are modelled with Voronoi cells and the high burnup structure is established through small isolated particles. Five tests conducted in the FR2 program were chosen to validate the simulation results, in which the predicted missing fuel lengths fitted well with the experimental observation, indicating the high fidelity of the simulation framework. Based on the validated simulation results, the mass fraction and filling ratio of fuel fragments along the axial direction were discussed to investigate the effects of cladding hoop strain on fuel relocation. Finally, a correlation of maximum mass fraction of fuel fragments with respect to cladding hoop strain was proposed to provide an implication to high burnup fuel performance evaluation under loss-of-coolant accidents for the future work.
Assumption of the Framework for the Evaluation of the Actual Effectiveness of Nuclear Safety Public Communication
Authors: Yang Lili , Xing Dan , Yang Yanfei , Chang Meng , Gong Yu 
Published: November 2022

Abstract
It is well known that Not In My Back Yard and public acceptance issues have become important factors affecting the development of the nuclear industry. Public communication is considered to be an important tool to address these issues. Both government agencies and operating organizations have invested considerable resources in public communication efforts. However, whether these investments have achieved their intended purpose is a question to be evaluated. Previously, there have been peer review of public communication efforts, but the focus has been on how the operating organizations do their jobs. The evaluation envisioned in this paper focuses on whether communication is effective, such as whether the Not In My Back Yard has been mitigating and public acceptance has increasing. In general, the evaluation framework envisioned in this paper breaks down “effectiveness” into different dimensions,divides different assessment areas based on the components of public communication efforts, distinguishes between daily works and important point-in-time activities, and focuses on the effectiveness of communication tools and the acceptability of content.In terms of assessment methods, they can include questionnaires, interviews, media reports, cited data analysis, etc., with more emphasis on using data external to the operating organizations. When conditions permit, regular public acceptance surveys can be conducted on a relatively fixed sample population to observe long-term trends. The evaluation results can be used to guide the subsequent public communication efforts, adjust the focus of work in a timely manner, and allocate resources appropriately.
Dynamic Modeling and Characteristic Analysis of Helium Brayton Cycle System
Authors: Liao Shengyong , Qiu Leilei , Bi Xiaolong , Sun Peiwei , Wei Xinyu 
Published: November 2022

Abstract
The Brayton cycle using helium gas as a working medium has the advantages of high conversion efficiency, simple design, compact structure and high inherent safety. It can be combined with fossil energy, nuclear energy, solar energy and other forms of heat sources, and has broad prospects for development. The Helium Brayton cycle (HBC) is a highly integrated system. The dynamic characteristics of the Brayton cycle system are complex and difficult to obtain. It is necessary to establish the mechanism model of the HBC for research and analysis, and provide the basis for subsequent control system design. Based on the equations of conservation of mass, energy, momentum and energy balance, the thermo-hydraulic models of heat accumulator, precooler, intercooler and pipeline are established in this paper. The model was established according to the characteristic curves of helium turbine, low-pressure compressor and high-pressure compressor, and the shaft speed model was established according to the energy balance equation. Coupled with the above models, a lumped parameter nonlinear model of the HBC is established. Based on the MATLAB/Simulink platform, typical transient conditions such as external heat source power disturbance and external load disturbance are calculated. The results show that with the decrease of heat source input power, the temperature of each equipment and the net efficiency of the circulation decrease. When the external load fluctuates, the imbalance of compressor and turbine torque will lead to the rapid change of shaft speed.