bluuebunny/crossref_metadata_2025_split huggingface dataset:
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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.