bluuebunny/crossref_metadata_2025_split huggingface dataset:
Thermal Cut-Off Energy for Accurate Analysis of Prismatic High-Temperature Gas-Cooled Reactor
Authors: Takeda Satoshi , Kitada Takanori , Yamamoto Akio , Yamaji Kazuya , Koike Hiroki , Asano Koji 
Published: November 2024

Abstract
In this study, the appropriate thermal cut-off energy for neutron transport calculations in High-Temperature Gas-cooled Reactor (HTGR) is investigated. The research employs the Monte Carlo calculation code MVP3.0 to evaluate the impact of changing thermal cut-off energy on the multiplication factor in the single fuel element geometries, focusing on variations in fuel and moderator temperatures at both the Beginning Of Life (BOL) and End Of Life (EOL). Up-scattering effect mainly caused by graphite used in HTGR as moderator has strong relation to the thermal cut-off energy. The findings show that when the thermal cut-off energy is set at 20 eV, the multiplication factor tends to be underestimated, especially at high fuel temperature. Furthermore, the study confirms that when the thermal cut-off energy is at or above 30 eV the variations in multiplication factors remain within approximately ±50 pcm.
An Iterational Physical Decoupling Method for Assessing the Safety of Deformed Fuel Channels
Authors: Zheng Junzheng , Lin Musen , Yang Xingtuan , Wang Dingqu , Li Songyang , Hao Wentao , Xiong Wei , Jiang Yueyuan 
Published: November 2024

Abstract
Irradiation during reactor operation causes continuous deformation in the fuel channels, resulting in narrowing of the control rod’s channels, which directly affects the safe insertion of the control rods. Faced with a multi-physics coupling scenario where the control rod is affected by flow resistance and channel resistance at the same time. This paper proposes a calculation scheme based on physical field decoupling and continuous iteration to calculate the rod dropping process in the deformed channels. Specifically, firstly, based on the empirical formula of the flow resistance, substitute the initial solution to estimate the flow resistance, and apply it to the control rod in the finite element method. Secondly, the finite element method is used to estimate the interaction force between the control rod and the channels, and then combining the flow resistance to obtain the time-velocity relationship (new solution). Thirdly, determine whether the iteration is terminated or return to the first step. Our study provides a solution to evaluate the safety of rod’s dropping process.
Study on Parameters Design Methodology for Rolling and Pitching Test of Control Rod Hydraulic Drive System
Authors: Li Yanlin , Qin Benke , Bo Hanliang 
Published: November 2024

Abstract
Control rods are the most important power regulation components in nuclear reactors. The startup, shutdown, and the reactivity regulation of reactors can be achieved by adjusting the position of control rods. The function of control rod drive mechanism (CRDM) is to adjust the position of control rods. Therefore, the reliability of the CRDM is important to the safety operation of reactors. According to different installation locations of the CRDM, it can be divided into two types, as the built-in type CRDM and the built-out type CRDM. Compared with the built-out type CRDM, when the built-in type CRDM is used by reactors, the volume of the reactors can be reduced and the rod ejection accidents can be avoided. Based on the actual requirements of the small modular reactors (SMRs), the built-in type control rod hydraulic drive system (CRHDS) is designed by Institute of Nuclear and New Energy Technology (INET), Tsinghua University. The CRHDS is mainly made up of six components, as the combination valve, the spring box, the rod position indicator, the hydraulic drive mechanism, the control rods, and the buffer. The CRHDS can be used by many kinds of SMRs, including the 5 MW and 200 MW nuclear heating reactors which are designed by Tsinghua University. In the field of naval architecture and marine engineering, the status of SMRs is irreplaceable. Compared with land-based nuclear power plants, when reactors are used by the marine nuclear power platforms, the responses and the characteristics of the fluid and some key equipment in the reactors should be studied in detail under rolling and pitching conditions, for the aims that the ocean environment suitability of the reactors can be evaluated. For the CRHDS, a series of rolling and pitching tests are planned by INET, aiming to verify the performance of the CRHDS under the rolling conditions that are defined by the relevant standards. In the rolling and pitching test, the structure of the CRHDS can be equivalently represented by a straight line, with the length of 5801 mm. It should be noted that the rolling center is located at the bottom point of the line in the rolling test. However, the actual rolling center does not coincide with the rolling center of the rolling test. Therefore, based on the locations of the actual rolling center and the rolling center of the rolling test, the input parameters of the rolling test should be designed, aiming to ensure that the load at the peril point of the CRHDS provided by the rolling test can reflect the conditions defined by the standards. In this paper, the load analysis of the CRHDS under rolling conditions is conducted. The test parameters are designed based on the equivalent overload coefficients. The parameter combinations used by the rolling and pitching test are determined. The results shown in this paper can provide a theoretical guidance for the rolling and pitching test of the CRHDS.
Communication Requirement Analysis of Intelligent Operation and Maintenance System for Mobile Small Modular Reactor
Authors: Yu Fangxiaozhi , Dong Hao 
Published: November 2024

Abstract
The potential applications of small modular reactors include supplying power to remote mining sites and communities. These application scenarios make it put forward higher requirements for “unattended monitoring and duty” capability. In order to achieve its operation and maintenance technical requirements, compared with traditional large commercial pressurized water reactors, small modular reactors need to adopt advanced operation and maintenance technologies with higher intelligence level. This paper analyzes the application scenario of the “intelligent operation and maintenance system” of small modular reactors, and clarifies the requirements of the monitoring center of small modular reactors. It lays the foundation for the “subsystem feasibility analysis,” “subsystem requirement analysis” and “system overall technical scheme design” in the follow-up “scheme design and requirement analysis” stage.
Analysis of SMR Reactor Coolant System in APROS
Authors: Ye Zhu , Xinbo Wang , Xianwei Liao , Zhiyun Cai , Minghao Liu 
Published: November 2024

Abstract
The reactor coolant system (RCS) is the primary system of the nuclear power plant primary cycle. Some small module reactors (SMR) applied once-through steam generator and integrated reactor vessel. The characteristic of these reactors differs widely from the tradition nuclear power plant. APROS Nuclear simulation and analysis software, with the Comprehensive plant model covering reactor island, turbine island, electrical and automation systems, has been successfully used in a series of major nuclear power plant projects, including power upgrade, modernization, safety improvement, and new plant projects. A reactor coolant system model of SMR that contains both process and control elements is developed and simulated using APROS. Steady states analysis is conducted under different power load and both force/natural cycle. Analysis of Transient condition including increasing/decreasing power load and force/natural circulation switching is also conducted to acquire dynamic thermal hydrate parameter. Reference and guidance can be provided for system optimization and safety analysis.
A Mass Flow Rate Control System in the Novel Superheated Steam Supply System of NHR200-II
Authors: Li Zongyang , Hao Wentao , Zhang Wenwen , Li Weihua , Yang Xingtuan 
Published: November 2024

Abstract
The previous version of the low-temperature nuclear heating reactor, rated at 200 MW (NHR200-II), was primarily dedicated to winter heating, hence limiting its applicability. Furthermore, the previous version of the steam generator within the secondary circuit was vertical and capable only of producing saturated steam. To expand its functionality, a novel superheated steam supply system has been integrated into the secondary circuit to generate superheated steam for an industrial park. This novel system comprises three crucial components: a preheater, a horizontal steam generator, and a superheater. The horizontal steam generator features effective fouling discharge, allowing the coolant in the secondary circuit to be softened water rather than the deionized water used in the previous vertical steam generator. NHR200-II comprises the primary, intermediate, and secondary circuits. The primary circuit operates via natural circulation under all operating conditions, accompanied by two passive residual heat removal systems designed to extract decay heat after reactor shutdown. These passive residual heat removal systems also rely on natural circulation, with reactor decay heat serving as the heat source and air-cooling towers acting as the cooling mechanism. However, the intermediate and secondary circuits are driven by pumps, generating superheated steam for industrial park use. The pressure in the intermediate circuit is slightly higher than that in the primary circuit, ensuring that no radioactivity would be released into the secondary circuit. The entire system is simulated by using RELAP5 code, achieving rated power conditions by employing design values. The heat transfer between tube and shell sides for the three important heat exchanger is modeled using the ‘Heat structure’, while a ‘Separator’ model is utilized atop the horizontal steam generator to separate water and vapor. Reactor core power adjustments are performed via the point reactor model, modified by introducing positive or negative reactivity through a logical control system. Additionally, a control system of the feedwater is implemented to regulate the mass flow rate within the secondary circuit, aiding in achieving different steady states with changing operational conditions. Simulation results demonstrate the effectiveness of the feedwater control system in efficiently adjusting the mass flow rate of the secondary circuit. Consequently, the entire system can attain a different steady state following alterations in operational conditions. Moreover, the temperature of the secondary circuit’s superheated steam exceeds 30°C, with a void fraction reaches to 1.00, meeting the specified requirements of the industrial park. The secondary circuit’s mass flow rate aptly corresponds to variations in reactor core power, ensuring the establishment of a new stable operational state.
Dynamic Modeling of Modular High Temperature Gas-Cooled Reactor Cogeneration System and Control Design Under the Condition of Turbine Isolation
Authors: Lin Xuan , Dong Zhe , Chen Fan , Zhou Shuqiao , Guo Chao , Li Duo , Sun Weidong 
Published: November 2024

Abstract
The HTR-PM600s modular high temperature gas-cooled reactor (mHTGR) has the advantages of inherent safety characteristics and high steam quality, and its application to combined industrial heat supply and electric power cogeneration can improve its economy. However, when the turbine is isolated, the modules must perform load adjustment to adapt to user’s side power changes, and the turbine’s main steam regulating valve will be unable to participate in the pressure control of the steam generator. In order to design an automatic control system to ensure stable operation and load following ability of heat island in the case of turbine isolation, a dynamic simulation model based on MATLAB/Simulink and Simscape toolbox is firstly established. The steady state simulation tests at typical operation condition points and the dynamic response experiments were carried out to verify the rationality and accuracy of the model. Based on the established model, the coordinated control scheme for NSSS under the turbine isolation condition is designed, which regulates the feedwater pumps and valves at the steam generators’ inlet. Simulation results show that the control scheme and control law can ensure the safe and stable operation of the cogeneration plant, it also can realize the load following of heat island without the turbine.
The Prospects of Small Modular Reactor Development and Application in China Based on the User Requirements
Authors: Zhu Sheng , Yan Zhen 
Published: November 2024

Abstract
The use of nuclear energy is one of the greatest discoveries of the twentieth century. After more than half a century of development, nuclear energy has become an important part of the world’s energy system. Small Modular Reactors (SMR) have the characteristics of small size, light weight and strong application flexibility, and can be used in various scenarios such as power generation, heating and seawater desalination.
Since the beginning of the 21st century in China, in response to the needs of heating in cold areas, power supply in remote areas and seawater desalination in offshore areas, major domestic nuclear energy enterprises and universities have developed various types of small reactors with different power levels and technical characteristics.
The development and application of small modular reactors should first consider the needs of potential users and should be sufficiently competitive in terms of nuclear safety and economics. From the perspective of technical maturity, industrial manufacturing base and current actual needs, small PWRs have obvious advantages, especially some advanced integrated small PWR models have relatively perfect engineering implementation conditions, and can be applied in a short period of time. However, metal cooling reactors and steam cooled reactors (except HTR-PM) are still in the stage of technology research and development, and some key technologies have not yet been fully broken through, and long-term technical research is still needed.
In this paper, the applicable scenarios of various types of small modular reactors from the perspective of user needs based on the characteristics of different technology types are summarized, and put forward the need to continuously optimize the design scheme of small reactors and enhance their market competitiveness, which can provide a reference for the development and application of small modular reactors.
Analysis of Coordinated Development Between SMR and Papermaking Industry in China
Authors: Ping Wang 
Published: November 2024

Abstract
The world’s first onshore commercial SMR, the Linglong-1 ACP100 in China began construction in July 2021. The ACP100 SMR has the characteristics of high safety, modularity and multi-purpose. SMR can provide steam for the industrial department.
The papermaking industry in China is a typical energy-consuming industry, which mainly relies on traditional fossil energy for steam and power supply. China is a major producer and consumer of paper. In the context of global warming, the industry is facing important issues in optimizing the energy structure.
The paper industry usually has its own small coal power plants and boiler houses to provide the steam and electricity required by the factory. SMR can be deployed to the user, which can effectively replace the papermaking factory’s own coal-fired power plant and coal-fired boiler.
From the perspectives of technical characteristics, technology maturity, decarbonization demand, and coal-fired power plant substitution, the coordinated development of China’s paper industry and ACP100 SMR is a typical case of nuclear energy helping industrial decarbonization. In this paper, the parameter characteristics and capabilities of ACP100 SMR in the field of industrial steam supply were analyzed. The development status, distribution characteristics and steam parameters of China’s papermaking industry are introduced. The coupling development opportunities between ACP100 and the paper industry were studied. The suggestion for SMR in the decarbonization transformation of papermaking industry is given.
AWCC Simulations Based on Monte Carlo Code RMC
Authors: Gou Yuanhao , Liu Zhaoyuan , Jia Conglong , Zhang Dacai , Luo Hao , Wang Kan 
Published: November 2024

Abstract
Neutron multiplicity measurement technology utilizes the correlation of neutron emission times during fission in nuclear materials. By employing mathematical tools to interpret processes such as neutron generation, multiplication, and detection in nuclear materials, a point model equation can be derived through this process. Simultaneously, nuclear material mass can be inverted based on neutron multiplicity counting. Due to the different spontaneous fission neutron yields of uranium and plutonium, neutron multiplicity analysis methods are divided into active and passive methods. This article is based on the independently developed Monte Carlo program RMC, which models and analyzes the AWCC device, and studies the neutron multiplicity counting in both active and passive methods. At the same time, the parameters of the point model equations are calculated, and the influence of different gate widths on the results is analyzed, demonstrating the capability of RMC in solving this problem. This expands the application of RMC in this field and provides reference for the design of subsequent experimental instruments.
Conduct a Tritium Radiation Safety Assessment on the Heating System Directly Supplied by the Steam From the Secondary Loop of the High-Temperature Gas-Cooled Reactor
Authors: Wang Wenyi , Li Chuan , Cao Jianzhu , Xie Feng , Li Hong 
Published: November 2024

Abstract
With the increasingly urgent demand for industrial heating, traditional heating methods are no longer suitable for the current energy development situation. The High Temperature Gas-cooled Reactor (HTGR), as a fourth-generation advanced nuclear energy technology, has several advantages, including good inherent safety, lack of pollution, no greenhouse gas generation, as well as its unique high quality steam generation and process heat applications. These attributes will significantly contribute to the strategic transformation of industrial parks and the expansion of industrial chains. Rational, safe, and effective use of nuclear energy is an important issue for China’s future new energy development. Nuclear heating can not only meet the demand for district heating and improve the quality of life of residents, but also implement a green and low-carbon energy development strategy and help achieve the “dual carbon” goal.
The current design of nuclear heating systems can be categorized into two types of schemes according to the source of the steam: the first is to use the steam in third loop for heating, which has the advantage of better safety due to the low level of radioactivity concentration in the third loop; the second is to use the steam in secondary loop directly, which can significantly improve the quality of heat supply and ultimately improve economy. No matter what type of HTGR process heat application, the influence of Tritium must be considered. Tritium is a main radioactive source item in the secondary loop of the HTGRs, if we try to use the HTGR secondary loop directly, Tritium will affect the users with much higher concentrations than third loop scheme, and the radiation safety needs to be carefully assessed. In order to maximize the utilization of heat energy, this paper studies the second type of heating systems and designs the tritium activity concentration in the supplied steam. The total emission of tritium in the radioactive effluent and the internal exposure dose of tritium for people under extreme conditions were calculated. All the assumptions of the calculations comply with the conservative principle of radiation safety design.
Finally, the feasibility of the scheme is discussed by comparing the calculated results with the limit requirement specified in the national standard. The results of the study can guide industrial design and provide a theoretical basis for the heating system directly supplied by the steam from the secondary loop.
Economical Research on Hydrogen Production of Sodium-Cooled Fast Reactor
Authors: Dai Zhiwen 
Published: November 2024

Abstract
The Sodium-cooled fast reactor could use the reactor fuel more efficiently and produce less long-lived radioactive waste, which will play a key role in World’s carbon peaking and carbon neutralization goal in the future. Owing to the high temperature at the reactor’s outlet (above 500°C), high-efficiency (above 40%), and high-tech maturity (more than 400 reactor years), which can provide the thermal power and electrical energy for hydrogen production simultaneously. This paper analyzes hydrogen production through CFR1000 using HEEP (V2021) software, which was developed by the International Atomic Energy Agency (IAEA). The results show that the most suitable hydrogen production methods for SFR are the copper-chlorine cycle and conventional electrolysis, considering the reactor’s outlet temperature and the goal of zero-carbon emission. The economic analysis shows that the cost of hydrogen production by water electrolysis through CFR1000 is 3.04 $/kg, through IFR is 2.84 $/kg. And this paper also conducted the hydrogen transportation method through the nuclear power plant. The research result shows that converting hydrogen into ammonia for off-site transportation is the best solution Considering safety and economy.