bluuebunny/crossref_metadata_2025 huggingface dataset:
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

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

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

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

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

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

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

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

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

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

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.
Seismic Fluid-Structure-Interaction Analysis of NHR200-II Fuel Assembly
Authors: Lin Musen , Yang Xingtuan , Wang Dingqu , Zheng Junzheng , Hao Yuchen , Li Songyang , Hao Wentao , Xiong Wei , Jiang Yueyuan
Published: November 2024

The fuel assembly (FA) stands as a pivotal component within the reactor core, impacting safety, reliability, and economic viability of the entire facility during the lifetime. As a newly-designed nuclear heating reactor, the FA of the 200 MW nuclear heating reactor (NHR200-II) embodies a novel configuration, characterized by a 9 × 9 fuel bundle arrangement inside, which are clamped by three spacer girds along the height direction, and fully enclosed by a square zirconium channel outside for improving effective coolant flow rate. Under the earthquake event, possible deformations of fuel cladding or outer channel, buckling instability of spacer grids, even damage of the structures and control rod insertion failures can occur and further pose unaffordable safety risks. Therefore, it is necessary to implement a comprehensive seismic fluid-structure-interaction (FSI) analysis with regard to the NHR200-II fuel assembly to evaluate the structural integrity.

In this paper, a particle finite element method (PFEM) based partitioned two-way FSI scheme is used to capture the seismic responses of NHR200-II FA, that is, implicit finite element method for structure dynamics, PFEM for fluid motion. To accurately solve the FSI phenomenon, it is essential to couple the fluid solver and structural solver advancing with a same time step, i.e., strongly-coupled on the FSI interfaces. The specific fluid solver and structural solver are combined together in the commercial finite element code LS-DYNA to carry out the two-way FSI simulation, which greatly improves computational efficiency. This finite-element based numerical framework has been validated in detail in the previous work on the axial flow-induced vibration analysis of a single NHR200-II fuel rod. At present, the seismic excitations applied to a fuel assembly level are not available, which are usually generated from several analyses of a site-specific earthquake, including sequential analyses of soil-structure-interaction model, the reactor pressure vessel (RPV) internals and FA array models. Thereby the computation employed a time-compressed El-Centro N-S earthquake (1.5 times amplification, ∼0.36 g peak acceleration) for the preliminary evaluation of the seismic FSI responses of NHR200-II FA.

The results present stress distribution of each component (including the top & bottom nozzle, fuel cladding, zirconium channel, spacer grids), channel-spacer grid interactions (transverse impact force), dynamic clamping of the spacer girds, acceleration responses, and response spectra of the FA at typical positions. The results would confirm the reasonable structural design and robust seismic performance of NHR200-II FA and also are broadly applicable to the seismic response of fully submerged fuel assembly with square-type channel outside.
Isotopic Measurement of HTR-10 Irradiated Fuels Based on Anti-Compton Gamma Spectroscopic Method
Authors: Zhang Weijian , Xiao Haiyan , Liang Jingang , Li Ruihan , Zhang Liguo
Published: November 2024

Accurate determination of isotopic compositions of irradiated fuel elements plays a critical role in assessing fuel burnup, optimizing reactor performance, and ensuring safe operations. Spherical fuel elements are used in pebble-bed HTGRs (High Temperature Gas-cooled Reactors) such as HTR-10 and HTR-PM. However, there is almost no isotopic measurement of such fuels in the open literature, which has seriously limited the optimization of fuel management practices for HTGR systems. Recently, an isotopic measurement experiment of two irradiated HTR-10 fuels is carried out in China using gamma spectroscopic method. The analysis results are presented in this paper.

We identified 9 fission products in the measured gamma spectra, and the measurement uncertainty are smaller than 5% for most isotopes. Furthermore, the correctness of the detection efficiency curve is stressed in this work. Multiple validation methods are employed, including standard source measurement, linear attenuation model, and self-calibration method, to ensure the accuracy and reliability of this work. Finally, sources of uncertainties are discussed and methods for accuracy improvement are proposed accordingly.
Research on the Dependency Between Transuranic Isotopes and Burnup in Pebble Bed High-Temperature Reactors
Authors: Hongjian Zhang , Qing Zhu , Liguo Zhang , Tao Ma
Published: November 2024

The analysis and measurement of transuranic isotopes in nuclear fuel constitute a crucial aspect of nuclear material accounting, holding significant implications for nuclear security and the prevention of nuclear proliferation. In traditional reactors, the analysis of nuclear fuel during operation can be conducted by inputting initial nuclide information and irradiation history into burnup simulation software for forward burnup calculation. However, in pebble bed high-temperature reactors, characterized by the use of numerous fuel spheres to form the core, the movement of these fuel spheres introduces uncertainties, making it impractical to track and record every individual fuel sphere. This characteristic renders the irradiation history of fuel spheres unknown, preventing the application of traditional nuclear material analysis methods.

Currently, the estimation of transuranic isotopes of fuel spheres in pebble bed high-temperature reactors is based on the correlation between transuranic isotopes and burnup. However, the accumulation of major transuranic isotopes has a strong dependency on irradiation history of spheres, resulting in that this correlation is not entirely accurate. Therefore, this paper undertakes work in the following two aspects. 1) Through an analysis of the flow characteristics of fuel spheres within the core, the paper summarizes the variation patterns in the irradiation history of fuel spheres. Utilizing the burnup chain of major heavy metal elements and burnup equations, it theoretically calculates the uncertainty of characterizing major transuranic isotopes using burnup indicators (such as Cs-137). The paper also analyzes the range of this uncertainty. 2) Based on a continuous diffusion model, the core was divided into channels and layers in conjunction with the neutron flux distribution in the pebble bed high-temperature reactor core. Using a random sampling method, a large number of irradiation histories for fuel spheres were generated. The study further investigated the correlation between major transuranic isotopes and burnup indicators using nuclear inventory calculation software.

The research findings of this paper demonstrate that the method for estimating transuranic isotopes in pebble bed high-temperature reactors, only based on burnup depth, exhibits a certain level of systematic error. To further enhance the precision of nuclear material accounting, it is imperative to establish new methods for nuclear material analysis. In addition to improving accounting precision, the outcomes of this research also contribute to establishing or improving a fundamental physical model for the irradiation history of fuel spheres. Furthermore, they can serve as an analytical foundation for benchmarking studies related to nuclear materials in pebble bed high-temperature reactors.
Neutronics–Mechanics Coupling for Fast Transient Simulation in Molten Salt Reactors
Authors: Vidril Théo , de Lambert Stanislas , Lelong Nicolas , Drui Florence , Patricot Cyril , Merle Elsa
Published: November 2024

Prompt supercritical transients initiated by high and rapid reactivity insertions are an important safety concern for fluid-fuel nuclear systems like molten salt reactors. Because of their short time scales, their study must involve a compressible treatment of the fluid fuel with finite velocity of mechanical waves. Otherwise, the delay in the negative reactivity feedback due to fuel expansion and the deformation of the vessel resulting from strong propagating pressure peaks would be overlooked. For a liquid fuel, this is especially true if it contains gas bubbles. Accordingly, and owing to the highly multiphysical nature of such excursions, a new neutronics/compressible-thermal-hydraulics numerical coupling tool based on the APOLLO3 and EUROPLEXUS codes is presented. An analytical solution to the coupled problem close to the one of the Nordheim-Fuchs model is also given. Excellent agreement between simulations and theory is obtained and demonstrates the successful implementation of the numerical tool.
Fluid-Structure-Soil Interaction Study for Nuclear Facilities With Large Pool Water Considering Nonlinear Structural Behavior
Authors: Sato Yuki , Ghiocel Dan M. , Kataoka Shunji , Morimoto Yasutomi
Published: November 2024

This study investigates the effects of the Fluid-Structure-Soil Interaction (FSSI) on the seismic responses of a deeply embedded nuclear facility with large water pools under severe earthquakes. This paper presents new seismic analysis method considering FSSI using 3D FEM for deeply embedded typical RC (Reinforced Concrete) shear wall nuclear buildings. In the new method, the commercial SSI analysis software, ACS SASSI was used with Option AA-F, which treats ANSYS fluid element FLUID80 dynamic matrices in ACS SASSI environment. By using the method it was confirmed that large pool water in the building provides significant local deformation of the pool walls, while the effects on other part of the structures are almost negligible from engineering point of view. To investigate more with the structure behavior during severe earthquakes, the nonlinear behavior of the structure was also taken into account with the new FSSI method using ACS SASSI NQA Option NON. To model the RC wall nonlinear behavior, their nonlinear backbone curves (BBCs) were computed using both Japanese and America standards. Then, the nonlinear analysis responses with both standards were compared. It was confirmed that the nonlinear structure behavior produced a visible shift of the ISRS spectral peak responses to the lower frequencies in comparison with the response of linear analysis. Also, although some significant differences of seismic responses were not observed between Japanese standard and American standard, it was confirmed that the differences of formulae in each code were reflected on their seismic responses.
Further Study of China’s Strategy for Closed Nuclear Fuel Cycle Development Against the Carbon Neutrality Target
Authors: Chen Ding , Shi Lei , Su Jiqiang , An Yan , Wang Yihan
Published: November 2024

Achieving carbon neutrality before 2060 represents a solemn commitment China has made to the global community. It is an arduous and challenging task for China to address climate change as a large developing country undergoing industrialization. In this endeavor, as a clean, safe, and efficient energy source with low carbon emissions, nuclear energy plays an important supporting role. According to estimates by Chinese and foreign research organizations, China’s installed nuclear capacity will reach an impressive 300–400 GWe by 2060, presenting substantial opportunities for development in this sector. However, China also faces significant challenges associated with ensuring a stable supply of natural uranium resources, safety management of spent fuel, and radioactive waste minimization, among others.

Over the past decades, China has steadfastly adhered to a closed nuclear fuel cycle policy, achieving significant milestones in technological research and development, engineering design, and equipment manufacturing. This study provides a comprehensive review of China’s historical efforts, current state, and pertinent challenges in closed fuel cycle development. Drawing insights from the experiences of other nations, this study puts forward novel perspectives on China’s nuclear closed fuel cycle policy: (1) The closed nuclear fuel cycle represents the only and most viable pathway for China to achieve nuclear sustainability; (2) Urgent efforts are needed to develop spent fuel reprocessing capabilities, particularly for PWR; (3) Spent fuel reprocessing and recycling must be developed in a coordinated manner; and (4) Consistent policy support must be provided to promote closed nuclear fuel cycle development.
Comparison Analysis and Optimization Research of Open and Closed Air Brayton Cycle
Authors: Song Meihui , Leng Yuyang , Qian Yiran , Chen Weixiong , Yang Yi , Yan Junjie
Published: November 2024

Microreactors have a significant impact on the strategy of energy safety. Due to the significant advantages of high security, compactness, high thermo-electric conversion efficiency, easy acquirement and supplement on working medium and so on, the air Brayton cycle system can serve as a reliable and dispatchable power conversion system of microreactor mobile nuclear power source in the remote islands or polar regions. Air Brayton cycle configuration adapted to different environments has a broad application prospect, which leads to comparison analysis and optimization work on open and closed air Brayton cycle systems.

In the present study, a 5 MW thermal heat pipe microreactor combining a simple recuperated air Brayton cycle energy conversion system is designed for mobile power sources at remote sites. Using a system-component combined design method, the influences of open and closed cycle parameters on the system performance are investigated and compared. On this basis, optimization work is carried out by NSGA-II, in which cycle efficiency and system power density are considered. Finally, a comprehensive design-analysis-optimization strategy is established to adapt to different environmental assumptions.

The results indicate that at the same power level, open and closed Brayton cycle systems reveal respective advantages. The closed cycle system shows higher cycle efficiency and system power density than the open cycle system. The open cycle system has more compact structures due to the lack of the cooler and indicates less estimated cost due to mature and advanced techniques of components, which demonstrate its advantages compared to the closed cycle system.
Fundamental Experiment of Gas Entrainment Phenomenon From Free Liquid Surface in a Sodium-Cooled Fast Reactor
Authors: Hamelberg Jasmine , Kobayashi Shunsuke , Endo Kazuki , Takeda Jotaro , Sakai Takaaki
Published: November 2024

The Sodium-cooled Fast Reactor (SFR) is an advanced type of nuclear reactor using liquid sodium as a coolant. However, the structure of the reactor and the flow of the coolant causes Gas Entrainment phenomenon to occur on the surface of the liquid sodium in a reactor vessel. Gas Entrainment causes the inert cover gas, i.e., argon gas, to be trapped inside the reactor vessel which may lead to disturbance of the reactivity. Hence, to maintain safe and stable operations a solution to prevent Gas Entrainment is required. Currently, researchers are developing a method to predict Gas Entrainment based on experimental data. At Tokai University, a special experiment apparatus is used to conduct water experiments to simulate Gas Entrainment phenomenon from a moving vortex dimple on the free liquid surface. Measurements of the inlet velocity and downward velocity were done by using the Particle Image Velocity (PIV) method. Image data of the vortices were recorded with the highspeed video system and the vertical length of the gas core in vortices from the liquid surface were evaluated. In addition, the number of vortices occurred were counted and classify the types of Gas Entrainment. From analyzing these experimental data, we expect to determine the mechanism of Gas Entrainment and how it occurs. In case of the special experiment apparatus at Tokai University, the high downward velocity around the downstream region allows the vortices to extend towards the bottom of the flow channel.
Numerical Simulation of Helium-Xenon Gas Recuperator With Different PCHE Channel Configurations
Authors: Li Dong , Xie Linfeng , Peng Cheng , Zhang Ziyue
Published: November 2024

The Brayton cycle system utilizing helium-xenon gas as the working material is one of the most promising solutions for space thermoelectric conversion. Within the system, the high-temperature recuperator plays a pivotal role for power generation efficiency and system compactness. Printed circuit plate heat exchanger (PCHE) is considered as the type of compact recuperator with large surface area, while the internal channels of PCHE can be arranged by many approaches. In this study, numerical simulations are conducted to investigate the flow and heat transfer performance of helium-xenon gas mixtures within PCHE, analyzing the influence of different geometric channels in PCHE, including zigzag and wavy channels with different corner angles, various channel diameters. Based on results of overall friction coefficient, Nu, recuperator effectiveness and synergy angle, it is recommended for wavy channel with the angle of 100° and diameter of 2mm to be adopted.
Numerical Investigation on PBM Kernel Functions for the Internal Flow Field of a Steam Separator
Authors: Yue Ma , Qianfeng Liu , Huang Zhang , Lantao Luo
Published: November 2024

In a nuclear reactor, the saturated steam generated by the steam generator needs to be dried by a steam-water separator, otherwise the impact and deposition corrosion of the steam droplets on the turbine blades will bring serious safety risks and reduce efficiency. In this work, the computational fluid dynamics software ANSYS Fluent combined with the population balance model is used to study the internal flow field of the steam-water separator. In order to correctly solve the internal flow field of the steam separator with PBM model, it is necessary to select the breakup kernel function and the convergence kernel function suitable for the internal working condition of the steam separator. The model of Luo and Svendsen (1996), the model of Lehr et al. (2002), the model of Coulaloglou and Tavlarides (1977) and the model of Prince and Blanch (1990) were selected. The relevant UDF was compiled and imported into the computational fluid dynamics software ANSYS Fluent, coupled with the PBM model for calculation, to solve the change of the volume fraction distribution of droplet agglomeration and crushing in the flow field of the steam separator, and the experimental results were compared. The influence of different kernel functions on the calculation results was explored, and the accuracy of the calculation of the separation efficiency of the water-steam separator under different kernel functions was analyzed.
Experimental Investigation on the Gasification Kinetics of Irradiated A3-3 Matrix Graphite
Authors: Wang Weishuai , Zhang Naizhe , Liu Xuegang , Xie Feng
Published: November 2024

Thermal treatment in the atmosphere of carbon dioxide is proved to be a more promising approach to selectively separate C-14, which contributes substantially to the radioactivity of irradiated graphite even after thousands of years, from the bulk of irradiated graphite. Some key factors such as the origin of graphite, the type of reactor and the atmosphere during treatment play significant roles in the thermal treatment. In this paper, irradiated A3-3 graphite was selected to study the gasification reaction kinetics in a range of CO2/Ar/CO gas mixtures. In the absence of CO, the results of gasification reaction rate in different CO2 partial pressure and temperature were presented. In the CO2/Ar/CO mixed gas, the gasification reaction rate in different CO partial pressure was also examined. According to the results of experimental studies, the activation energy value and pre-exponential factor of the global reaction and elementary reactions were obtained. As a conclusion, it is inferred that uniform gas-solid reactions occur on the surface of irradiated A3-3 matrix graphite particles within the temperature range of 975 to 1045°C. The results might be beneficial for the future design of thermal treatment for the selective separation of C-14 from irradiated A3-3 matrix graphite.
Layers of Protection When Adapting Modern Technologies to Enable Remote Handling of Radioactive Materials
Authors: Syed Sohaib
Published: November 2024

Canadian Nuclear Laboratories (CNL) is planning to decommission the Waste Management Area (WMA) standpipe and bunker structures at the Whiteshell Laboratories (WL) site located in Pinawa, Manitoba. The decommissioning activities will include retrieving historical radioactive waste from standpipes and bunkers. The waste stored in standpipes and bunkers ranges in activity from Low-Level Waste (LLW) to Intermediate-Level Waste (ILW). The standpipes and bunkers include radioactive solid waste stored in containers of unknown integrity and radioactive liquid waste, a byproduct of water ingress into the waste storage structures. The unknown integrity of waste containers presents a high risk of worker exposure to radiological hazards.

In recent years, CNL has conducted extensive research developing automated systems for handling radioactive waste during decommissioning activities. Adopting modern technologies and tooling to reduce human interaction with radiological material minimizes worker exposure. The design concept of these systems is based on the operational experiences and lessons learned (internal and external) from the WL Hot Cell Facility.

When implementing modern technologies and tooling to handle radioactive materials remotely, it is important to implement defence-in-depth in the system design (multiple layers of protection) to ensure the safety of workers, the public, and the environment. These layers of protection act as barriers to prevent a Postulated Initiating Event (PIE) from progressing into a severe accident. The more independent the layers are, the more robust the defence.

Layers of protection in the design of a system include inherently safe design, control and monitoring of the system, alarms requiring operator intervention or critical alarms with safety interlock to prevent an accident, passive or active protection to mitigate the consequence of the hazardous event, and emergency response. The system designs were conservative, with safety margins between normal operations and abnormal conditions.

This paper outlines some of the recent efforts of CNL in radioactive waste management, particularly in designing systems for radioactive waste handling activities involving historic solid waste. These activities include retrieving, characterization, sorting, segregation, and repackaging of radioactive waste. A high-level hazard identification methodology is used to identify hazards and PIEs associated with the design and operation of these remotely operated waste retrieval and handling systems. These hazards and PIEs are assessed and compared with the non-automated retrieval method to demonstrate that remotely operated waste retrieval and handling systems reduce the risk of radiological exposure to workers and protect the environment.
Development of an Extra-Trees-Based Program to Calculate Gamma-Ray Exposure and Energy Absorption Buildup Factors
Authors: Chen Zhitao , Liu Yongkuo , Hu Jifeng
Published: November 2024

The calculation of gamma-ray energy absorption and exposure buildup factors is studied in this paper. The GP formula requires logarithmic interpolation calculation about the photon energy and the equivalent atomic number of the material in the interpolation calculation of the fitting parameters, which leads to a complicated process of calculation of the buildup factor. In order to solve this problem, a program for buildup factor calculation based on the Extra-trees regression algorithm has been developed in this work, which can calculate the energy absorption and exposure buildup factors by inputting the photon energy and the elemental composition of the material. The code was validated across a wide range of materials by calculating the buildup factors for a variety of mixtures/compounds, including water, air, concrete, human tissues, and bricks. The tests show that the results of the program are in good agreement with the values from the ANS, GP formulas, and MCNP program. Moreover, the accuracy of the Extra-trees regression model is improved in this paper by using the GP formula to generate the buildup factors of elements 4–92 as a training set compared to using ANS data.
Multi-Mesh Approach for Geometry-Independent Neutronic/Thermal-Hydraulics Coupling Analyses Using RMC and Ansys Fluent
Authors: Zhao Xingyu , Liu Guodong , Huang Shanfang , Chen Qiaoyan , Luo Hao , He Ying , Hou Junren , Wang Kan
Published: November 2024

This paper presented a multi-mesh approach for high-fidelity and geometry-independent neutronic/thermal-hydraulics (N-TH) coupling calculation capable of various nuclear reactor analyses using Reactor Monte Carlo code (RMC) and Computational Fluid Dynamics (CFD) code Ansys Fluent. Verifications and coupling analyses were performed on various models.

Customized RMC/Fluent interfaces were developed for direct data transfer through hdf5 files without text files to achieve higher transfer efficiencies and precisions. This study presented a multi-mesh approach to realize mesh mapping between solvers by separating the data statistical Q-mesh and T-mesh from the computational cells. Q-mesh tallied the power distributions in 3D structured forms using mesh tallies of RMC, while T-mesh tallied the temperatures and densities in Fluent model by the UDFs. Both solvers read data through Q-mesh and T-mesh without directly interacting with each other, and thus the geometry-independent coupling was realized. This study also considered relaxation method during iterations to stabilize coupling solutions and improve convergence.

Verifications on a unit cell model showed great agreement to previous studies with a maximum fuel temperature deviation less than 0.6%. Coupling calculations performed on a light water-cooled unit cell model and a helium-cooled assembly model both converged in 3 iterations. The results showed that the coupling effects were dominated by the moderator density variation. The maximum fuel temperature of coupled results was 17.6% higher than that of decoupled results, and 7% higher than that of the coarse model, showing the conservative nature of high-fidelity coupling, and thus the high-fidelity coupling was beneficial for reactor safety.
The Effective Thermal Conductivity Model of Dispersion Fuel Elements Based on Effective Matrix Approximation
Authors: Qiu Tianchen , Sun Jun
Published: November 2024

The modular high temperature gas-cooled reactors use fuel elements with huge number of Tri-structural isotropic (TRISO) fuel particles dispersed in graphite matrix, which contributes to the inherent safety and enables high temperature process heat applications. Featured by complex geometry and material combinations, a lumped model of effective thermal conductivity (ETC) is practically used to describe the heat transfer performance and temperature distribution of the fuel element the in thermal hydraulic design and safety analysis of HTGRs. Since the fuel particles release fission or decay heat, classical analytic ETC models fail in predicting the temperature distribution inside the fuel elements, without denoting the inner heat sources. Previous researches have explained the definition of ETC model with inner heat sources. Some new ETC models that indicate the function of distributed inner heat sources in the dispersion fuel element were also proposed.

In this paper, an ETC model to predict the volumetric average temperature of dispersion nuclear fuel elements is proposed. The proposed model considered the influence of distributed inner heat sources in the composite media by introducing effective matrix approximation. The ideal dispersion fuel pebble, which is simplified from the element currently used in pebble bed HTGRs, is taken as an example to show the application of the proposed ETC model. And the validity and conservativeness of the proposed ETC model are preliminarily revealed by finite element numerical calculation results.
A Preconditioning Method Based on Strongly Implicit Procedure for the Multi- Diagonal Equation
Authors: Tan Bo , Zhang Haojie , Wen Yutong , She Ding
Published: November 2024

Ellipsoidal partial differential equation is usually discretized into a multi-diagonal equation when using the finite-difference or finite-volume numerical method. The neutron diffusion equation in reactor physics and the thermal conduction convection equation in thermal fluid calculations are elliptical partial differential equations. Therefore, the efficient solution of multi-diagonal equation is of great significance to nuclear engineering calculations. GMRES algorithm has been widely used to solve multi-diagonal equations, but the computational efficiency of this method depends on the effect of preconditioning method. This paper combines the SIP preconditioning method with GMRES to solve steady-state thermal conductivity problems for flat cartesian and cylinder. The calculation results show that the SIP preconditioning method can reduce the iterations of multi-diagonal equations, without increasing the additional preconditioner’s memory compared with the traditional ILU(0) preconditioning method.