bluuebunny/crossref_metadata_2025 huggingface dataset:
Affecting Factors and Aspects of Developing Nuclear Public Acceptance Strategies
Authors: Lin Haomiao
Published: November 2022

Nowadays, the issue of climate change is moved from margins to the mainstream. The important role of nuclear power in the global fight against climate change has been significantly highlighted by IAEA at COP26 last year. As a low carbon base-load electricity provider, nuclear power can became progressively an important option of countries for achieving its sustainable development goals. Comparing with traditional energy supply, nuclear power projects are characterized by complex technology, long construction period, large capital demand and high sensitivity. Currently, the international nuclear power demand market is mainly concentrated in the emerging countries, most of which have a poor industrial infrastructure, “not in my backyard” phenomenon is prominent, which is seriously interrupting the government’s energy development decision. Different national conditions and development phases will also have different impacts on public perception. This paper focuses on the analysis of the common characteristics of factors affecting public acceptance in different markets, and puts forward suggestions on public communication strategies, so as to provide references for improving public acceptance.
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

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

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

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

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

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

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

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

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

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

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

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

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

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.
Study on Multi-Objective Optimization Design of LP Section for 1000MW Class Nuclear Turbines
Authors: Hu Yifeng , Hu Zhiqiang
Published: November 2022

With the development of nuclear reactors, the domestic market demand for high-capacity nuclear power turbines is also growing. Low pressure (LP) section (include LP inner casing and LP outer casing) is one of the important components of nuclear steam turbine, which are mainly welded by plates and rods. The increasing size of LP section leads more vacuum load, more self-weight, more stress, more deformation and so on. Therefore, the traditional design methods can no longer adapt to this situation very well.

In this paper, the stress, deformation, buckling behaviors and seismic loadcase of the LP section of steam turbine are evaluated by the numerical finite element analysis. Then, the multi-objective optimization (include size optimization and topology optimization) design of the LP section is done, so that the structural efficiency and performance of LP section is achieved. The performance of all aspects of the optimized LP section has been greatly improved. On the basis of a small improvement in stress and deformation, the self-weight of the LP section is greatly reduced. At the same time, the optimization boundaries, the choice of the model and the process are discussed. This study proposes a more optimized design method of LP section of steam turbines, and it offers considerable economic benefits.
Application of Risk-Informed Approach in Emergency Compressor Emergency Backup Function Test
Authors: Liu Deyi , Zha Sixing , Wu Jian , Cao Yong , Wang Zilong
Published: November 2022

The emergency compressed air production system is a part of the compressed air production system of the nuclear power plant. Its function is to provide the necessary compressed air to the nuclear island when the main air compressor fails and cannot provide compressed air. Each unit of the nuclear power plant has two emergency air compressors to provide backup air source for the compressed air distribution system for instrumentation. Under normal working conditions, one emergency air compressor is in the basic load state and the other is in the standby state. The two emergency air compressors will automatically start according to the degree of pressure drop of the compressed air production system. The purpose of the emergency backup function of the emergency air compressor is to verify that the two air compressors will start up as the system pressure drops in different states, and to verify the operating parameters of the two emergency air compressors. The test period is one week. Periodic testing is an important activity for nuclear power plants to ensure the availability of equipment that performs important functions. However, periodic testing exceeding a certain frequency may reduce the safety of the power plant and increase the unnecessary burden of the power plant. At present, PSA has been widely used in nuclear power plant operation guidance, maintenance rules and many other fields, PSA has developed into an important tool for safety assessment and decision-making. With the development of PSA technology, the risk-oriented model combining probability theory and determinism has also been applied in the optimization of the periodic test cycle of nuclear power plants, so that the resource allocation of power plants can be optimized under the premise of ensuring sufficient safety of nuclear power plants. Improve regulatory and operational efficiency. This paper will take the emergency compressed air production system of Fangjiashan Nuclear Power Plant (FSJ NPP) as an example to introduce the application method of the risk-oriented periodic test optimization method.
Comparative Machine Learning Study for Estimating Peak Cladding Temperature in AP1000 Under LOFW
Authors: Najar Merouane , Wang He
Published: November 2022

For a more realistic estimation of safety margins, the conservative approach is replaced by integrating the best estimate approach (BE) with uncertainty quantification, the integration which knows as best estimate plus uncertainty (BEPU), which can predict the key safety parameters such as peak cladding temperature (PCT) and departure from nucleate boiling ratio (DNBR), etc. In this sense, a fast and cost-effective tool for uncertainty quantification is developed through a data-driven approach to predict PCT under loss of feedwater accident (LOFW) in AP1000 reactor.

This paper includes performing a comparative study between different regression ML algorithms to find the best algorithm which can predict the PCT with higher accuracy.

Intent to generate the required data for training and testing the ML algorithm, an uncertainty quantification framework is developed by coupling a best estimate code (RELAP5) with a statistical tool (RAVEN). RELAP5 is used to simulate the thermal-hydraulic response under LOFW accident while a set of uncertainty parameters are propagated through the RELAP5 model using RAVEN. These distributions were sampled using a Latin Hypercube Sampling (LHS) technique to generate sets of sample cases to simulate using the RELAP5 code. 5,000 runs were generated in order to acquire a large database for training purposes. The examined algorithms are linear regression, supported vector machine, k-nearest neighbors (KNN), and random forest. The evaluation of algorithms depends mainly on mean absolute error (MAE) and determination coefficient R2.

The result shows that the random forest provides high accuracy in predicting PCT within four algorithms, which reaches 98.96%.
Heating Start-Up Method and Demonstration for Full Natural Circulation Reactor
Authors: Yu Jianhui , Tan Wenji , Zhang Xingliang
Published: November 2022

Since there is no main coolant pump (MCP), the full natural circulation reactor can not use the MCP to heat the reactor coolant system to normal operating temperature prior to reactor criticality. This paper presents an external starting heating method. The principle and brief system description of the external start-up heating system is presented. By reasonably arranging the inlet and outlet nozzles of start-up heating in the reactor, the external start-up heating source is used to heat the reactor coolant system, by establishing natural circulation in the reactor.

In order to demonstrate the feasibility of external start-up heating system, the heating effect of start-up is analyzed, and the analysis results are verified by experiments. On the other hand, the impact of external start-up heating system on reactor operation and safety is evaluated. In order to ensure that the external start-up heating system does not affect the operation and safety of the reactor, the suggestions on the configuration and operation of external start-up heating system are given.

By the external start-up heating system, the problem of startup heating without main pump of full natural circulation reactor is successfully solved. Therefore, it provides an option of hot start for the full natural circulation reactor.
Assessment on Accident and Mitigation of Spent Fuel Pool in a Typical PWR
Authors: Guo Chao , Ge Zhenzhen , Liao Feiye , Chen Peng , He Dongyu
Published: November 2022

In consideration of low power rate and large water inventory in spent fuel pool (SFP), the accident progression may be very slow and the fuel melt accident was considered to be very unlikely in SFP. However, the accident in Fukushima Dai-ichi Nuclear Power Plant implies the possibility of severe accident in SFP and highlights the importance of accident management of SFP. As the fuel building has no three-barriers containing radioactive products like the reactor building and SFP has significantly larger fuel inventory, the potential consequences of severe accident (radioactive materials release and hydrogen production) could be far beyond the capability of practical mitigation measures in design. Since Fukushima accident, lots of efforts related to SFP simulation are made by developers of several severe accident simulation codes. For ASTEC code, models about radiative heat transfer and Zircaloy oxidation are added in V2.1 which makes it possible to have a better prediction of severe accident in SFP. In this paper, representative scenarios (loss of coolant, loss of cooling) in SFP of a typical pressurized water reactor (PWR) design are simulated with ASTEC code. Accident progression and phenomena are presented. The influence of potential mitigation measures is studied. Suggestions to the design of SFP and accident mitigation strategies are provided according to the analysis results.
Research on Fretting Wear of Grid and Cladding Tube Under Simulated Primary Circuit Environment of Pressurized Water Reactor
Authors: Liu Tingting , Li Kun , Ma Wenhui , Zhang Yuxiang , Chen Minli , Wu Xiaohang
Published: November 2022

With the deepening of the fretting wear process, the wear behavior and wear mechanism of the contact surface between the grid and the fuel rod cladding will gradually change, but there are relatively few research reports in this area at home and abroad. In this paper, the chemical environment of high temperature and high pressure water in the primary circuit of the PWR is simulated. In this environment, the friction pairs formed by the grid and the cladding tube are subjected to fretting wear tests of different durations (192h, 288h, 384h). The cladding samples is subjected to wear scar detection, and the wear topography data (including the wear volume and maximum wear depth) of the contact point between the cladding tube and the grid clamping system are obtained, and the wear coefficient is finally calculated using the ARCHARD equation. The results show that with the increase of the test time, the spring wear scar data (including the wear volume and the maximum wear depth) formed by the grid spring and the cladding tube tend to increase, the wear rate is basically stable at 0.8μm3*s−1, and the wear coefficient is basically stable at 0.4 × 10−17 Pa−1, there is no obvious change. For the dimple wear scar formed by the dimple of grid and the cladding tube, the wear scar data (including the wear volume, the maximum wear depth and the wear coefficient) are significantly greater than those obtained under the test conditions of 192h and 288h at 384h. However, the data obtained at192h is slightly larger than that at 288h, combining with the Archard equation, it can be seen that this may be caused by the decrease in the wear rate and wear coefficient in the middle of the fretting wear. The results of this paper show that with the deepening of the fretting wear process, the wear characteristics of the grid and cladding samples are constantly changing, it has a good reference significance for the existing fretting wear research between the grid and the cladding tube and the related nuclear fuel structure.
Study on High Frequency Flow Induced Vibration Mechanism and Evaluation Method of PWR Fuel Assembly Grid
Authors: Ding Jie , Zhang Chaozhu , Sun Jingya , Gan Fujun , Gong Shuhe
Published: November 2022

By the analysis of the three-dimensional flow field of the fuel assembly grid, the high-frequency flow-induced vibration test of small-scale test assembly, the preliminary research on the overall and local flow-induced vibration mechanism and influencing factors of the fuel assembly grid is carried out. A system of evaluation index based on the three-dimensional flow field analysis of the grid and the flow-induced vibration test of 5 × 5 small test assembly is preliminarily established, which can more accurately evaluate the flow-induced vibration characteristics of the PWR fuel assembly grid.
Working Fluids Analysis and Discussions of the Stirling Cycle for Space Nuclear Reactors
Authors: Yang Chenhao , Zhuang Nailiang , Zhao Hangbin , Tang Xiaobin
Published: November 2022

Space nuclear power is the most potential power source for future deep space exploration, interstellar navigation, and planetary surface base station. For medium and high-power space missions, an efficient, compact and reliable energy conversion system that converts nuclear reactor thermal energy into mechanical energy or electrical energy is critical for the entire power system. The Stirling cycle that converts thermal energy into mechanical energy through expansion and compression of working fluids has the advantages of strong load adaptation, high conversion efficiency, modular combination, robust reliability, redundancy, etc. Therefore, it is very suitable for future high-power deep space missions. The technical route of space nuclear power proposed by Nanjing University of Aeronautics and Astronautics is based on the modular Stirling thermoelectric conversion and liquid molten salt core scheme. The present study is the simulation prediction of the thermodynamic performance of the Stirling engine. The working fluid’s composition, physical properties, and leakage properties have great influences on the thermal efficiency, operating lifetime, and reliability of the Stirling cycle. In this regard, a comprehensive study on the properties and effects of H2, He, He-Xe mixture, N2 and air as working fluids was carried out in the present study. The modified Stirling thermodynamic model IPD-MSM was used to simulate the Stirling cycle, and the heat and power losses caused by various irreversible factors were analyzed. Moreover, impacts of operating pressure, heat source temperature, and piston frequency on every heat and power loss were discussed. The results show that the pressure loss and the non-ideal heat transfer loss are the dominant losses. However, different working fluids and operating conditions have different performances of power loss. The effect of operating frequency and working pressure are significant, while the effect of operating temperature is relatively small. The present study provides theoretical support for selecting thermoelectric conversion methods for future medium and high-power space missions.
Evaluation of Detectability of Pump/Diagrid Link Rupture in Pool-Type Sodium-Cooled Fast Reactor
Authors: Onoda Yuichi , Uchita Masato , Tokizaki Minako , Okazaki Hitoshi
Published: November 2022

The safety analyses were carried out to confirm the sufficiency of the function of the plant protection system against the pump/diagrid link rupture. The target plant is a pool-type SFR of about 600 MWe class equipped with an axially homogeneous core currently under development in Japan. In the pool-type SFR, the primary system piping connects primary pump and the high-pressure sodium plenum located at the inlet of fuel sub-assemblies and called “pump/diagrid link”. Because this piping is submerged in the reactor vessel, it is difficult to detect small scale sodium leakage in this piping, and thus a certain large pipe break like guillotine should be assumed and evaluated as a design basis event. In order to confirm the detectability of pump/diagrid link rupture by safety protection system signals, a series of analyses of the guillotine break for a pump/diagrid link were carried out. Sensitivity study had also been performed to consider the uncertainty of the reactivity coefficient in the analyses. The sufficiency of the function of the plant protection system against the pump/diagrid link rupture was confirmed by the analysis results that at least two signals are transmitted for the detection of the event, which is the development target of the plant protection system in pool-type SFR.
Investigation on the Active Learning Optimization on Rotor Dynamics of SCO2 Turbine
Authors: Zheng Zhaoli , Wu Jun , Ke Zhiwu , Zhao Zhenxing , Yang Xiaohu , Dai Lu
Published: November 2022

The supercritical carbon dioxide (SCO2) Brayton cycle has the characteristics of high power density and high thermal efficiency, which is an important development direction of the micro power plant. SCO2 turbine is the core component of the SCO2 Brayton cycle of which the dynamics have important influences on operational reliability of the entire system. With regard to the rotor of SCO2 turbine, the equations of motion is established by adopting finite element method. The complex eigenvalues of the rotor are solved in the state space, and the campbell diagram has been drawn to obtain the critical speeds. The steady state response of the rotor is obtained by the harmonic balance method, and the safety of the system is estimated based on API684. Results show that the rotor is not safe with its original geometric paramters. Aiming to improve the operational safety, an optimization method based on active learning is developed to maximize the separation margin. Results show that after the optimization, the separation margin has been greatly increased. Comparing with the genetic algorithm (GA) and the parttern search (PS) method, the iteration number of the active learning optimization method has been greatly reduced. The effectiveness of the developed optimization method is proved, and the study method and conclusions can serve as a reference to the optimization of SCO2 turbine rotors in the industry.
Characteristic Behavior of Flow Field in Blocked Fuel Assembly Flow Channel under Steady Flow Condition
Authors: Jayeola Mathew , Tan Sichao , Qiao Shouxu , Ala Ayodeji , Dione Fulgence
Published: November 2022

Past histories of fuel assembly blockage incidents had made investigation of flow channel blockage in rod bundle a compelling field of thermal-hydraulic safety. This study experimentally investigate flow field structural characteristics in a 5 × 5 rod bundle under two sub-channel blockages with geometries; 2 × 5 and 2 × 3, in a non-mixing spacer grid. The blockage effects on velocity distribution and turbulence characteristics is examined under laminar, transitional and turbulent flow regimes across a range of Reynolds number between 1108 ∼ 8103. Under laminar flow, result of spatiotemporal analysis for flow downstream of the blocked spacer grid-SP2, showed formation of a large symmetric vortex pair shedding in wake formed at both edges of the blockage. For transitional and turbulent flow, the vortex gradually dissipate especially towards the flow channel’s wall due to interference of friction effect. Damping effect of blockage on upward velocity flow and pressure in the vortex dominated region creates a backflow zones accompanied by flow reversal. Analysis of turbulence characteristics examined for 2 × 5 blockage across adjacent bundle gaps and sub-channels showed immense variations with Reynolds number and channel wall impact. The study aimed to provide relevant reference basis to compare and analyze similar scenario under unsteady flow condition.
Experimental Study on the Relationship Between Performance and Structural Parameters of Metal Fiber
Authors: Li Lintai , Zhou Yanmin , Li Yuanyuan , Sun Zhongning , Wang Yin , Gu Haifeng , Ma Song
Published: November 2022

As a new filter material, metal fiber has good high-temperature resistance, moisture resistance, and radiation resistance. To explore the influence of the fiber’s primary structural parameters on the fiber’s performance, this paper experimentally studied the relationship between the flow resistance, filtration efficiency, and dust holding capacity of the metal fiber filter material and the thickness and solid volume fraction (SVF). Under the condition of ensuring uniform deposition of NaCl aerosol, we found that the deposition law of NaCl polydisperse aerosol is the same as that of monodisperse aerosol. The results show that for the same fiber diameter, increasing the fiber’s thickness and SVF will increase the flow resistance of the fiber under the same dust holding capacity; that is, reducing the dust holding capacity of the fiber. In the depth filtration, the efficiency increases rapidly due to the aerosol deposition in the fiber, while the efficiency changes gently when the filter cake is covered on the filtration surface. At the same time, for each high filtration efficiency fiber, the variation characteristics of resistance with dust holding capacity in the deep filtration stage are the same. After the filter cake is completely covered on the filter surface, the growth trend of resistance is the same, independent of the fiber’s structural parameters.
Study on Frequency and Time Domain Analysis of Flow-Induced Vibration Response of Core Barrel
Authors: Cao Guitao , Zhu He , Liu Guangdong
Published: November 2022

Different calculation methods of flow-induced vibration of the reactor vessel internals (RVI) are studied in this paper, that is frequency domain analysis and time domain analysis. Firstly, considering the fluid-structure interaction between the core barrel (CB) and the reactor pressure vessel (RPV), the CB and the heavy reflector, the finite element model of the CB in water is established and the results of its vibration characteristics are obtained, which is consistent with the test results. Based on the flow-induced vibration test of the RVI scale model, the fluctuating pressure data of the CB is obtained. The parameters such as power spectral density (PSD), correlation length and coherence function are obtained by processing the test data. Thus, the cross-power spectral density between any two points is got through these parameters. The root mean square (RMS) response of the CB is obtained by random vibration analysis (frequency domain). Then, the CFD model is established and large eddy simulation (LES) is used to obtain the time history of the fluctuating pressure of the CB. The response of the CB is obtained by time integration method (time domain). The calculation results by using these two methods are in good agreement with the experimental results.
ISO 19443: Fundamental Requirements of the New Quality Management Standard for the International Nuclear Supply Chain
Authors: Bargsten Jeanne , Treier Clemens , Yilmaz Özlem , Pades Amandine , Kim Andrea , Lee Jung-Kyu
Published: November 2022

The new quality management standard ISO 19443 [2] defines specific requirements for supply chain organizations in the nuclear industry. The standard became increasingly popular in the last years and a certification is now often a prerequisite for equipment manufacturers and service providers to participate in the bidding process of new build projects.

These companies face the challenge, to adapt their existing quality management system to the new requirements of ISO 19443 [2], since this is necessary to receive a globally accepted accredited certification.

A conducted analysis of the ISO 19443 standard [2] identified major differences to the established quality management system standard ISO 9001 [3], the results are presented. Furthermore, the key requirement areas of the standard with particular significance to the nuclear industry, are evaluated.

ISO 19443 [2] enshrines nuclear safety culture as the fundamental element throughout the entire supply chain and enables a graded approach to the application of quality requirements. The standard aims at the global harmonization of management standards and facilitates the regulation of the international nuclear supply chain.

Best practice approaches are highlighted, on how to successfully adapt a quality management system. It is based on experience of one of the first accredited inspection bodies, already supporting several manufacturing companies on their way to certification.