bluuebunny/crossref_metadata_2025 huggingface dataset:
Flow Regime Visualization of Wet Steam Flow in a Horizontal Circular Pipes Having Different Diameter
Authors: Uchiyama Yuta , Morita Ryo
Published: November 2024

To evaluate the occurrence conditions and rate of progression of flow-induced aging degradation phenomena such as pipe wall thinning and vibration fatigue in steam piping systems, it is essential to accurately understand the steam flow conditions inside pipes. However, the flow structure (flow pattern and sound velocity) of water-steam two-phase flow under high steam quality conditions is not sufficiently clear at present, and the occurrence and progression of degradation phenomena may not be properly evaluated. In this study, the pattern of water-steam two-phase flow under high-steam-quality conditions was visually observed to clarify the flow pattern. The database of flow patterns under a wide range of conditions was constructed by expanding the data, especially in the annular dispersed flow region for systems with pipe inner diameters of about 40 mm, and by acquiring new flow visualization data for piping systems with pipe inner diameters larger than 40 mm. It was shown that the existing flow regime maps can be used to determine the flow pattern of wet steam flow through comparison with existing flow regime maps of water-air two-phase flow, thereby improving the reliability of the flow pattern determination method.
Assessment of CFD Methods in Simulating Flowfield in a Vane-Type Separator
Authors: Zeng Xiaobo , Xu Yifan , Fan Guangming , Yan Changqi
Published: November 2024

The vane-type separator is an important device to remove the fission gas from molten fuel in Molten salt reactor. To better understanding its working mechanism and provide a reliable reference for its practical application, numerical study was carried out in this paper. Due to the centrifugal force, the transition of flow pattern happens, leading to the change in interface structure. Before bunching to the gas core, the scale of interface is less than millimeter, while for the scale of interface between gas core and continuous phase is much larger than that. Because of this multi-scale interface phenomenon, it is important to assess the CFD models to obtain a high-fidelity result. Therefore, two multiphase models were compared. One is the Euler-Euler method and another is a hybrid approach of Euler-Euler and Volume of Fraction methods. Besides, Large Eddy Simulation (LES) was used to capture the turbulence information of the flowfield. The result shows that a sharper interface between gas core and liquid is able to simulate using the hybrid method because of the introduction of the transportation equation of volume fraction.
Dynamic Response of Nuclear Fuel Transport Package Under Multiple Impact Angles
Authors: Hao Yuchen , Xu Bin , Zhang Xiaohong , Sun Wenzhao , Li Yulin , Liu Guang , Ma Yan , Li Yue , Lin Musen
Published: November 2024

Nuclear power is one of the most efficient way to provide the clean and green electrical energy. As the uranium in Earth is sufficient to meet global energy needs for several hundred years, it is a sustainable energy source. Meanwhile it can generate power 24 hours a day, unlike renewable energy sources that are rather dependent on weather conditions. To develop the nuclear energy better, the safety performance of nuclear energy cannot be ignored where the importance of safely transporting nuclear fuel has received increased attention. Nuclear fuel includes radioactive material that can pose a significant health threat to human and environment if not transported properly. Therefore, as nuclear transportation packages need to avoid radioactive material leakage in public domain, 9-m drop test is serious but necessary.

Although extensive research has been carried out on the most severe condition under the 9-m drop test, the main challenge of the actual test faced by many reseachers is the impact angle which is directly influenced by the weather and operating conditions. For example, the wind cannot be ignored in the actual because it can affect the impact angle which make it more difficult to predict the orientation of the nuclear fuel transport package. Therefore, this study aims to compare the dynamic responses under multiple impact angles where the typical nuclear transport package is chosen as the research objective. First the geometry model of the nuclear fuel transport package is introduced. The possible impact angles caused by the accidental factors are discussed and then used as the boundary condition in the 9-m drop analysis model. The finite element method (FEM) is used to solve the drop analysis model and then obtain the dynamic responses of nuclear fuel transport package via explicit non-linear software. Finally, the analysis and comparison based on the numerical result is carried out. It is hoped that this research will contribute to a deeper understanding of the influence of impact angle caused by the accidental factors in the actual test. Furthermore, this research approach can be applied in the design and validation of nuclear fuel transport package, which could make the package more robust and reliable.
Advanced Waste Monitoring: Field Trials of a Wireless Sensor Network for Radioactive Waste Package Integrity Assessment
Authors: Chierici Andrea , Ciolini Riccardo , Lo Frano Rosa , d’Errico Francesco
Published: November 2024

Radioactive waste materials, produced from nuclear technologies and power generation, must be safely contained and confined into suitable packaging systems that align with international global standards for disposal. One of the most important aspects related to package disposal is the possibility to monitor and control the physical and thermo-mechanical properties of the overall waste form over time. With recent technological advancements, there is now the capability to deploy flexible and distributed wireless sensor networks for long-term monitoring, diminishing reliance on intricate cabling infrastructure or mechanical scanning systems. This approach can substantially reduce inconsistencies and human errors that typically arise during the information management of packages throughout the disposal process, and integrating these technologies with decommissioning machinery, such as conveyor belts, robotic arms, and specific demolition equipment, may provide real-time radiation data during all stages of the decommissioning process. This study focused on validating and testing a novel system for identifying and continuously monitoring the integrity of stored radioactive waste packages developed at the University of Pisa. It is based on a network of wireless node devices equipped with solid-state, micro-power radiation detectors. Preliminary field trials were conducted at the Nucleco storage site in Rome, demonstrating the system’s efficiency in providing detailed insights into the activity distribution within cemented overpacks and the system’s ability to transmit data correctly even in harsh environmental conditions. Further, a joint demonstration test at ÚJV Řež, in line with the objectives of the H2020 European Project PREDIS, was conducted to emphasize the system’s ability to offer a comprehensive waste monitoring solution. The system’s novel ability to infer package deterioration and degradation based on activity data in a completely automated way sets it apart from existing methods, marking a significant advancement in destructive and non-destructive testing techniques.
Investigations on the Debris Bed Formation Behavior of the Heated Solid Particles
Authors: Zou Wenbin , Liu Zhaoran , Tong Lili , Cao Xuewu
Published: November 2024

During the later stages of the postulated Core Disruptive Accidents (CDAs) of a sodium-cooled fast reactor, the porous debris bed is expected to form in the lower plenum region. The geometries of the debris beds are significant for the coolability and the risk of neutronic subcriticality. In this paper, the sedimentation and formation behavior at different factors of the debris bed is experimentally investigated; the mound dimple diameter and mound dimple depth are chosen as the critical quantitative parameters, which can express the flatness of the debris bed intuitively. The results show that the particles formed a concave debris bed at normal temperature. The general tendency of developing a relatively large dimple area is observed with larger nozzle size, lower particle diameter, lower particle density, and higher water level. The prediction model of the mound dimple volume of the debris bed is established, and the error is less than 30%. When the particles are in the film boiling regime, the vapor film reduces the friction between the particle and fluid, thereby increasing the impact momentum and driving the debris bed in a flat direction.
Hydrodynamics of High-Density Particles in a Multistage Fluidized Bed With Downcomers
Authors: Zhang Yiming , Yang Ying , Wang Huan , Li Shuyue , Zhang Yongmin , Wu Xiuhua
Published: November 2024

Fluidized bed reactor is commonly used in uranium recycling processes of nuclear industry, particles used in these reactors usually have very high densities, which are usually several times higher than those of catalyst particles in chemical industry or fuel particles in fluidized bed boiler in power plants. Multistage fluidized bed reactor has advantages of better residence time distribution and higher gas-solids contact efficiency. However, less literature has involved the operation of multistage fluidized beds with high-density particles. In this study, we used a super-high-density powder of wolfram carbide (WC) powder, which had a true density of 15630 kg/m3 and a typical light powder of similar size, FCC particles, to investigate systematically the different hydrodynamics in a cold-model counter-current three-stage fluidized bed with downcomers. The column diameter, stage height, and downcomer diameter were 186mm, 400mm and 30mm, respectively. The ranges of superficial gas velocity and volumetric solids feeding rate were 0–0.2 m/s and 4–8 ml/s. Pressure drop, pressure fluctuation and solids flow rate at the bottom outlet were measured during stable operation status to evaluate the operating stability. The stable operating ranges were also determined in experiment. The experimental results indicated that this high-density WC powder could be fluidized and circulated continuously in this multistage fluidized bed, but lower fluidization quality was observed using WC than FCC. The minimum operating gas velocities of WC were higher than those of FCC particles, which was determined by the change of transient pressure signals. Different flow phenomena were observed at the minimum operating velocities using the two powders of different densities. The experimental results demonstrated the feasibility of operating uranium compound powders in multistage fluidized beds.
Analytical Study of Matching of Dual Compressor Modules With Turbine for Small Modular Reactors
Authors: Lu Frank K. , Jayamani Ananthkumar , Kim Daejong , Bolin John , Curiac Radu , Choi Hangbok
Published: November 2024

An analytical study of a power conversion unit consisting of a dual compressor and a turbine operating on a closed Brayton cycle with helium is presented. The turbomachinery was a constant hub design with two compressor modules and a turbine module. The design code included the capability to perform sweeps of input parameters such as the polytropic efficiency or loss coefficient, degree of reaction, etc. It is required that the pressure rise of each of the two compressor modules be approximately equal. An optimization procedure for the low- and high-pressure compressors resulted in both of them having seven stages. The performance for each of the compressor stages is almost identical. Another optimization procedure for the turbine yielded six stages. Likewise, the performance for each of the turbine stages is almost identical. Performance maps were developed individually for each of the compressor modules and the turbine. Mathematical constraints based on problem physics were used in conjunction with the component maps to trace turbomachinery operating points. Results show that the low-pressure compressor operates close to the surge line. It is recommended to implement a recirculation path to avoid possible surge in compressor.
Using Radius of Gyration in Order to Determine Surrogate Mechanical Properties of a Porous Structure
Authors: Shiman Oksana V. , Gaudet Michel
Published: November 2024

Advances in additive manufacturing have now made printing very fine design features possible, giving rise to the possibility of incorporating a porous structure within a component to act as a thermal barrier. The model of such a component may be simplified during finite element analysis by replacing the topographically complex porous structure with that of a homogeneous material having surrogate material properties equivalent to those of the porous structure. Determining changes in length within a porous structure for the purpose of calculating strain presents a challenge. Due to the nature of the porous structure, non-uniform displacements are possible, introducing difficulty in assessing aggregate displacements. The possibility of using the radius of gyration as a measure of displacements within a structure is explored in the present study. The positions of finite element meshing nodes were used to calculate mass distribution within unit cells of a gyroid structure, and from this, a radius of gyration was derived. Unit strain, and consequently Young’s modulus and Poisson’s ratio were obtained for the porous structure from the changes in radii of gyration, which can be used to describe an equivalent bulk material. Excellent agreement was obtained when directly comparing the elastic response of a gyroid cluster and equivalent homogeneous material under identical loading conditions.
Study on Power Regulation Methods of Microreactor Coupled With Closed Helium Gas Turbine Direct Cycle
Authors: Geng Xuyao , Wang Jie
Published: November 2024

Microreactor is a class of rapidly developing small advanced reactor. It has many advantages: mobility and transport ability, resilience, and independence from the grid, as well as its capacity for long refueling intervals. The microreactor utilizing high-temperature gas-cooled reactor technology has inherent safety. By coupling microreactor with the helium turbine direct cycle power generation system, it can operate at high temperature. Therefore, the microreactor coupled with closed helium gas turbine direct cycle (MR-GT) has broad prospects for development. In this paper, the mathematical model of the MR-GT system is established to study the dynamic characteristics. Based on the conservation equations of mass, momentum and energy, the thermo-hydraulic model of power conversion system including heat transfer components and turbomachinery is established. The transient reactor power is calculated by the point reactor neutron kinetics equations. Finally, the lumped parameter nonlinear system model of the MR-GT is established by coupling the above models. The Modelica simulation language is used to solve these nonlinear ordinary differential and algebraic equations. The dynamic characteristics of the MR-GT system under the conditions of reactivity insertion and helium inventory regulation are studied. The results show that the system power can be independently adjusted within a certain range by changing the reactivity or helium inventory. By synergistically regulating the reactor reactivity and helium inventory, the temperature and efficiency of the MR-GT system can be kept basically unchanged, which is an ideal power regulation method.
Numerical Investigation on the Thermal-Hydraulic and FIV Characteristics in LBE-Cooled Helical Cruciform Fuel and Wire-Wrapped Fuel
Authors: Zhang Qi , Wang Haoyu , Cao Junxian , Li Yuanming , Li Chenxi , Liu Zhenhai
Published: November 2024

The lead bismuth eutectic (LBE) cooled fast reactor in one of the General 4 advanced reactors, which plays significant role in the nuclear fuel cycle. In the recent years, the wire-wrapped fuel with four wires surrounds the cylindrical fuel is developed and promoted, which may be accept by the LBE-cooled fast reactor. The helical cruciform fuel of the Lightbridge corporation has similar geometrical features with the wire-wrapped fuel. The cross-section of helical cruciform fuel consists of the convex blades and the concave elbows, and the cross-section is rotated around the axis. The blade and elbow of helical cruciform fuel function as the wires while they are part of the nuclear fuel. In the present work, the comparison on the thermal-hydraulic and flow-induced vibration (FIV) characteristics of the helical cruciform fuel and wire-wrapped fuel is carried out by numerical simulation. The fluid-structure interaction is solved by the coupling between the fluid and solid domains, and the transient turbulence fluctuation is solved by large eddy simulation (LES). The results show that the convective heat transfer is enhanced by the spiral structures even though the hot spot phenomenon is caused simultaneously. The hot spot temperature of the helical cruciform fuel is larger than the wire-wrapped fuel duo to the non-uniform distributed heat flux. As the coolant in the flow channel is blocked by the blades and wires, the directional transverse flow is generated and thus the transverse mixing is enhanced. The one-way mixing with opposite direction occurs in the two sides of the single rod. There are similarities in the FIV characteristics for the spiral shaped fuels. The maximum vibration displacement and dominant vibration frequency of the helical cruciform fuel and wire-wrapped fuel are almost the same.
Development of Thermal-Hydraulic and Safety Analysis Code for a Heat Pipe Cooled Reactor
Authors: Jiao Guanghui , Xia Genglei , Zhou Tao , Wang Jianjun
Published: November 2024

Heat pipe cooled reactors have attracted attention as a promising option for nuclear power generation in space and deep-sea applications due to their simple design, scalability, safety and reliability. Although various models of heat pipe-cooled reactors have been proposed, there remains a gap in rapid and efficient thermal-hydraulic programs for evaluating the operational characteristics and safety of these reactors. In this study, a transient neutron-thermal coupling code was developed to assess the transient characteristics of the core of a heat pipe cooled reactor. This code includes a neutron physics model, a heat pipe model, and a core heat transfer model, and it was verified using reference data and experimental results. Additionally, the code was employed to analyze the operational characteristics of typical accidents, such as reactivity insertion accidents, heat pipe failure, and loss of heat sink incidents. In the reactivity insertion accident, the reactor power stabilized to a new steady state after a period. In the accident of heat pipe failure, the heat from the failed fuel assembly was effectively transferred through adjacent heat pipes. During a loss of heat sink accident, the reactor core safely shut down due to negative feedback, with an initial spike in core temperature followed by a decrease. Throughout all transient processes in these accidents, the fuel pin temperatures maintained a significant safety margin, demonstrating the reactor’s strong safety performance in various accident scenarios. This code thus provides an effective tool for the design and safety analysis of heat pipe cooled reactors.
Design and Experimental Verification of Passive Residual Heat Removal System for a Alkali-Metal Heat Pipe Cooled Reactor
Authors: Wu Hexin , Gou Junli , Wang Ruifeng , Yuan Leqi , Shan Jianqiang
Published: November 2024

The high-temperature heat pipe (HTHP) cooled reactor has the advantages of compact, safety and reliability, rendering it a promising technology for application in Unmanned Underwater Vehicles (UUVs). In this paper, a Passive Residual Heat Removal (PRHR) system is designed for a marine nuclear power plant proposed by the Chinese Academy of Engineering Physics (CAEP). The PRHR efficiently utilizes the space of the adiabatic section of the heat pipe bundle as Emergency Cooling Chamber (ECC) to establish natural circulation. The CFD model of the ECC was established and the geometric parameters of the system were optimized. The results indicate that a hexagonal baffle around the adiabatic section of heat pipe bundle is superior to no baffles. The inlet and outlet widths of the hexagonal baffle has a negligible impact on the heat removal power during the natural circulation process. Furthermore, the analysis shows that the residual heat removal power requirement can be conservatively met when the axial length of the ECC is 100.0 mm. Subsequent experimental investigations were conducted to analyze the characteristics of heat removal power and the wall temperature distribution of the heat pipe under various cooling conditions. The results reveal a logarithmic rise in heat removal capacity with the mass flow rate, a linear decline in cooling water temperature with increasing inlet temperature, and a negligible impact of pressure fluctuations on heat removal capacity. Furthermore, the mass flow rate, temperature, and pressure of the cooling water demonstrate minimal effects on the temperature of the heat pipe wall.
Comparative Study of Deep Learning Models for Accidents Classification in NPP: Emphasizing Transparency and Performance
Authors: Najar Merouane , Wang He
Published: November 2024

The nuclear power plant (NPP) plays a crucial role in providing clean energy, significantly contributing to mitigating global warming. However, this advantage is accompanied by public concerns regarding potential accidents. The imperative for a robust safety system has led to the exploration of artificial intelligence (AI) integration in NPP to minimize human error and enhance safety efficiency during reactor accidents. This research conducts a comparative study between two deep learning (DL) models to classify five accidents based on algorithm performance and transparency. These accidents include Loss of Coolant Accident (LOCA), Loss of On-site Power (LOOP), Loss of Flow Accident (LOFA), Steam Generator Tube Rupture (SGTR), and Feed Water Line Break (FWLB). The necessary data was collected using open data from Personal Computer Transient Analyzer (PCTRAN) software, subsequently separated into two categories: digital and numerical. Image data is utilized for training Convolutional Neural Network (CNN), while quantitative data is employed for training Artificial Neural Network (ANN). Additionally, four metric parameters (Accuracy, Recall, F1-score, Precision) are used to evaluate the model’s performance. Finally, eXplainable Artificial Intelligence (XAI) techniques are employed through the use of the Shapley Additive Explanation (SHAP) package to provide the model’s output explainability.

The results show that within the two DL algorithms, CNN presents the best performance, demonstrating an excellent accuracy reaching 99.74 and requiring less time and a smaller dataset than ANN. Furthermore, the use of XAI explains the reasons behind the models’ decision-making process, leading to the establishment of trust in the algorithms’ results.
Advancement in Shielding Materials: Geant4 Simulation of Gamma Radiation With Polystyrene-Bismuth Oxide
Authors: Aminu Bashir Garba , Liu Yongkuo , Akhdar Hanan , Lawal Shamsuddeen
Published: November 2024

The selection of suitable materials for radiation shielding is of paramount importance in safeguarding individuals and the environment against the detrimental effects of ionizing radiation. As radiation-emitting technologies continue to expand across various domains, the meticulous assessment of appropriate shielding materials has become increasingly critical. This work has presented a comprehensive study on Polystyrene-bismuth oxide composites (PS-Bi2O3) within an energy range spanning from 100 to 2000 keV, containing different percentage compositions of bismuth oxide (Bi2O3). The primary objective is to explore the gamma mass attenuation μ/ρ and linear attenuation μ coefficients of these composites which lead to comprehensive evaluation of half-value layers of all the studied samples as it is very crucial parameter in finding the shielding ability of a material, while it encompassed various weight factors of bismuth oxide within the samples. This exploration has been carried out through the utilization of EpiXS and a Monte Carlo simulation-based toolkit, namely Geant4. The rationale for covering such a wide energy spectrum for gamma rays is to ensure applicability across a diverse range of scenarios, so that the findings can be useful in many different situations. These attenuation coefficients have been used to determine the half-value layers of each of the various samples. Additionally, the study involved a comparative analysis of the effective atomic numbers and effective electron densities obtained from the EpiXS results.

In summary, this research offered valuable insights by not only developing new and reliable shielding materials but also by thoroughly exploring the shielding characteristics of Polystyrene-bismuth oxide composites across a diverse range of energy levels and examining potential of its uses across that wide range of energy levels. The findings have the potential to inform the selection of effective shielding materials in various practical applications, ensuring enhanced radiation protection in different settings.
Multi-Objective Optimization Study of Nozzle Outlet Size Based on Atomization Performance
Authors: Qingshan Chen , Wang Mingjun , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: November 2024

The nozzle plays a significant role in the nuclear power plant’s safety and the reactor’s integrity. The structure of the nozzle directly affects the atomization performance, consequently impacting the pressure and temperature variations within the regulator. This study focuses on the single inlet pressure-swirl nozzle structure, employing the response surface optimization method to obtain the influence of nozzle outlet diameter, outlet length, and outlet chamfer on atomization performance. The optimization responses encompass droplet size, droplet uniformity, and atomizing cone angle, and Design-Expert software is utilized to optimize the design of the nozzle outlet size. Compared to the pre-optimization state, the optimized nozzle exhibited a 43.44% reduction in average droplet diameter, a 30.74% increase in atomization cone angle, and a 76.73% improvement in droplet uniformity.
Overview of Tritium-Resistant Performance of Heat Transfer Tube Materials in Reactors
Authors: Wu Mengjie , Zhou Ziling , Wang Yu , Qu Zhengzhe , Chen Jing , Xie Feng , Cao Jianzhu , Tong Jiejuan
Published: November 2024

Tritium, a radioactive hydrogen isotope widely distributed in fission and fusion reactors, has received considerable attention for its behavior within nuclear systems. As the sole barrier against tritium permeation from the primary to the secondary loop, the performance of heat transfer tubes in resisting tritium is extremely crucial. In future fusion reactors, where tritium serves as a fuel at higher concentrations and larger scales, its permeation issues will become even more challenging. Therefore, evaluating and developing heat transfer tube materials with high tritium resistance is paramount for advancing reactor high-temperature applications and ensuring their radiation safety.

This paper reviews the model of tritium permeation through heat transfer tubes from the primary to the secondary loop, using DT as an example. It then discusses the use and composition of common heat transfer tube materials in reactors, summarizes the tritium permeation rates, diffusion coefficients, and solubility at corresponding temperatures for different alloys, and analyzes the impact of Cr and Ni content on the tritium-resisting performance of alloys. After comparison, Hastelloy XR, F82H and Incoloy 800H are selected as three alloys with high tritium-resisting performance for use as reactor heat transfer tube materials. Finally, the paper provides feasible suggestions for preventing tritium permeation, hoping to offer valuable references for tritium control and management in both fission and fusion reactors.
Experimental Study of the Effect of Inclination Angle on the Heat Transfer Characteristics of Steam Condensation in the Vertical Tube
Authors: Liu Jiabao , Cao Xiaxin , Yang Peixun
Published: November 2024

The purpose of the experiment was to investigate the effect of inclination angle on steam condensation heat transfer characteristics for downward flow in a vertical tube at different mass fluxes. The inlet pressure of saturated steam was 0.2 MPa and the mass fluxes ranged from 20 to 60 kg·m−2·s−1. The experiment was conducted at a range of inclination angles between 0° to 25°. The results showed that with the variation of inclination angle, the flow patterns in the tube at the outlet of the condensate tube mainly showed three kinds of type, namely, annular wavy flow, semi-annular wavy flow and stratified wavy flow. Moreover, the heat transfer coefficient increased with the increase of the inclination angle, and the influence of the inclination angle on the heat transfer coefficient decreased with the increase of the mass fluxes. The effect of the inclination angle on the heat transfer coefficient increased as the vapor quality decreased. Comparing the heat transfer coefficients measured experimentally with the calculated values of the correlation equation of Shah (2016), the results showed that the precision of the predicted values of the relational equation of Shah for the heat transfer coefficients of condensation in the inclined tube was good enough.
Monte Carlo Simulation of Response of Beta-Delayed Neutron Inspection System for Fuel Cladding Damage
Authors: Zhang Weihua , Zhang Liguo , Cao Jianzhu
Published: November 2024

This work was aimed to design one novel device to monitor β-delayed neutron in the circuit coolant water. The response of β-delayed neutron inspection system was primarily simulated by PHITS Monte Carlo code including the pulse height spectrum, detection efficiency and time spectrum of neutron counters. The detection efficiencies (from 1 keV to 1 MeV) of two neutron monitors were respectively displayed as about 50% (12 He-3 counters) and about 30% (108 10B-coated counters). The beta monitor was designed as array of thin-slice plastic scintillators immersed in water to attain as large detection efficiency as possible. The simulated detection efficiency for beta particles (energy higher than 1 MeV) was more than 34%.
Temperature Field Rapid Estimation of Space Nuclear Thermionic Reactor Based on Reduced-Order Model
Authors: Liu Shuo , Wang Chenglong , Qiu Suizheng , Tian Wenxi , Su Guanghui
Published: November 2024

In order to overcome the shortcomings of Computational Fluid Dynamics (CFD), which needs to consume a lot of computational resources and time to obtain the flow field information. In this paper, a non-intrusive reduced-order model was developed based on the POD an BP neural network. Then the established ROM was applied in the temperature filed rapid estimation of the space thermionic reactor, TOPAZ-II. ANSYS Fluent are used to obtain the data snapshot samples of the temperature field of TOPAZ-II. Under the new operating conditions, the temperature field was rapidly reconstructed by training-completed ROM. The ROM was validated by comparing the ROM results with Fluent results. The maximum absolute error of temperature is 0.2 K with speed-up on the order 4.8 × 103. Therefore, the methodology in this paper provides a reference for real-time monitoring of the thermionic reactor operation in orbit.
Gas Leakage Test for Radioactive Materials Transport Package With Vibration
Authors: Sun Qian , Wang Pengyi , Chen Lei , Bai Juying , Jiao Limin , Wang Zhipeng , Zhuang Dajie , Sun Hongchao
Published: November 2024

Radioactive materials transport packages (RMTP) are the key equipments to ensure the safe transportation of radioactive materials, which can play a variety of important functions. Among them, the containment sealing function is an important safety requirement for RMTP containing radioactive contents such as gases and powders. The realization of the sealing for RMTP is not only to safeguard the quality of the contents and to prevent any reduction in the quantities of the contents. The leakage rate requirement should be determined at the design stage, which is as the criterion for judging the conformity of the sealing performance. And a suitable detection test method should be selected to carry out gas leakage test. Currently, gas leakage test for RMTP is mainly used to verify the sealing performance at the normal condition and the condition after tests required by IAEA SSR-6. However, the effects of different vibration conditions during transportation on the RMTP sealing performance are not taken into account. Therefore, the effects of vibration on the RMTP has been studied by experiment. Helium mass spectrometer leak detector and cladding inflation-gas detector method were used in the study. A scaled-down RMTP model and a small vibration test equipment were used carry out constant-frequency vibration tests. It was found that the cumulative effect of vibration causes the gas leakage rate increase for RMTP and it increases with the number of vibration tests.
Validation and Verification of ASYST Code for Predicting Condensation Phenomena in Nuclear Reactor Safety Systems
Authors: Saraswat Satya Prakash , Hassan Mubashir , Osman Sameer Mohammad , Allison Chris , Addad Yacine
Published: November 2024

The ASYST (Adaptive System Thermal-Hydraulic) code series is a recent development that integrates the functionalities of SCDAPSIM and SAMPSON. It can simulate and analyze fluid and reactor systems with Best Estimate Plus Uncertainty (BEPU) capability. The proposed work involves conducting a thorough validation and verification process for the ASYST code model to predict the condensation phenomena using a condensation-separate effect test facility (present at the Department of Mechanical and Nuclear Engineering, Khalifa University), specifically focusing on the condensation phase. The condensation separate effect test’s experimental facility was devised by members of the thermal-hydraulic team affiliated with the Nuclear Engineering Department at Khalifa University’s Emirates Nuclear Technology Centre. This apparatus is specifically configured to acquire dependable data on condensation heat transfer analysis in a vertical tube simulating the condensing element within a Passive Containment Cooling System (PCCS), wherein steam and non-condensable gas mixtures are examined. The aim is to provide a detailed description of how the ASYST code aligns with and accurately represents the outcomes of the test facility.

The condensation of the separate effect test facility is a critical benchmark for this validation and verification framework. By scrutinizing the code’s performance concerning the test facility outcomes, this approach comprehensively evaluates the ASYST code’s reliability and applicability to practical scenarios of nuclear reactors’ passive safety systems.
RCCA Ejection Accident Analysis Based on Coupled 3d Neutronics and Thermal-Hydraulics
Authors: Liye Zang
Published: November 2024

The rod cluster control assembly (RCCA) ejection accident is one of the most important Design Basis Condition (DBC) of Pressurized Water Reactor (PWR) nuclear power plant. This accident leads to a rapid reactivity transient by inducing an uncontrolled positive reactivity, followed by a nuclear power increase and a power distribution disturbance. The nuclear power increase is limited by reactivity feedback effects, such as the Doppler feedback effect caused by the fuel temperature increase. This may result in Departure from Nucleate Boiling (DNB) among the adjacent fuel channels, and the potential failure for the fuel cladding.

At present, the traditional decoupling analysis method is still used in RCCA ejection accident (REA) analysis of HPR1000, which has a high degree of design conservatism and restricts the continuous improvement of the unit’s economy and performance improvement. This paper takes a certain HPR1000 reactor as the research object, establishes a coupled 3D neutronics and thermal-hydraulics analysis method based on SMART/FLICA III-F code. The analysis process and results were compared with traditional method. The research results indicate that the coupled 3D neutronics and thermal-hydraulics analysis method is reasonable and feasible, which can more realistically simulate the core situation during the REA transient process, and can effectively release the conservatism of key parameters. On the basis of ensuring the same level of safety, it provides the possibility for improving the performance of the unit.
Simulation of Aerosol Transport Under Spray Control Measures for Floating Reactor Accidents
Authors: Liu Zhe , Wang Ning , Yang Ya-Peng
Published: November 2024

A three-dimensional accurate simulation was conducted using computational fluid dynamics methods for the spray removal of aerosols after a large break accident in a floating reactor. The Euler Euler Lagrange method was used to simulate gas solid liquid, and grid independence work was carried out to verify the reliability of the grid model. The physical model adopted takes into account the inertial collision, interception, and diffusion effects of droplets on aerosol particles. The simulation results can accurately provide the flow field and droplet motion path distribution in the cabin. At the same time, the removal effects of different mechanisms on aerosols, as well as the removal effects of single droplets and droplet groups on aerosols, are compared.
Investigation on Isotopic Depletion Validation and Uncertainty Analysis With Kalman Filtering
Authors: Qiao Jianshu , Huang Dongli
Published: November 2024

Isotopic concentration is one of the essential indicators to evaluate the burnup process, which provides detailed descriptions of fuel assembly status among various reactor core conditions. This work aims to develop a method in support of isotopic depletion validation. The developed validation method attempts to improve predictions of isotopic concentration across burnup fusing the computational results from a high-fidelity model with a small amount of measurement data. The isotopic predictions are improved by minimizing the discrepancies between the model prediction and the reality, as well as reducing the associated predicting uncertainties. However, this is such a challenge due to the lack of experimental measurement data and limitations by qualities of existing data. Isotopic validation method developed in this work takes advantages of Kalman Filtering (KF) to solve the challenges mentioned above. This work makes the best use of the reduced data to optimize the prediction process based on a few measurements across burnup. The simulations and uncertainty analysis are completed by OpenMC with a buffer code. Results show that the discrepancies between computational predictions and the reality are reduced, and uncertainties of isotopic concentrations across burnup decrease by about 50%.
Analysis of Energy Dissipation Effect of Fluid-Filled Capacitive Structure in the Water Hammer Process
Authors: Yang Linqing , Qin Benke , Li YanLin , Ma Yue , Bo Hanliang
Published: November 2024

Water hammer is a common phenomenon in nuclear engineering piping systems. Water hammer can cause a large fluctuation of fluid pressure, which may threaten system security. It is essential to control and mitigate pressure fluctuations and energy. The fluid-filled capacitive structure is a key component of energy dissipation equipment, and plays an important role in the water hammer mitigation process. The water hammer performance experiment rig was established, and the experiments were carried out. The water hammer theoretical model was established, and the influence of unsteady friction was considered. The model results agreed well with the experimental data. The fluid-filled capacitive structure was added to the pipeline system to analyze the dissipation effects using two analysis methods. The analysis results show that as the fluid-filled capacitive structure moves from upstream to downstream and the cavity volume increases, the fluctuation amplitude of pressure and internal energy decrease, the fluid mass transfer between the capacitive structure and the pipes is enhanced, and the energy dissipation effect increases. The capacitive structure periodically absorbs and releases the energy of the water hammer process to achieve the energy dissipation effect. The research results can provide guidance for the design of water hammer pressure mitigation and energy dissipation equipment in nuclear engineering pipeline systems.
CFD Analysis of Heat Transfer in Molten Salt Fuel Chambers of the Wielenga Innovation Static Salt Reactor (WISSR)
Authors: Khaleb Ikhwan , Wielenga Thomas , Yang Won Sik
Published: November 2024

This paper presents the computational fluid dynamics (CFD) analysis to study the heat transfer in the Wielenga Innovation Static Salt Reactor (WISSR), which is a fast-spectrum TRU-burner reactor. The core consists of thin annular fuel chambers holding NaCl-(TRU,U)Cl3 fuel salts that are arranged concentrically, and NaCl-KCl-MgCl2 coolant flowing through the gaps between fuel chambers removes the heat. In the central fuel chambers, stainless steel corrugations are used to enhance the conduction of heat from fuel to coolant. The CFD analysis focuses on two main aspects. Firstly, it assesses the natural circulation of fuel salt. Secondly, the interaction of fuel salt and helium gas at the top of chambers is investigated using the volume of fluid (VOF) method. The CFD analysis confirms the feasibility of the design.
Numerical Investigation of Postulated Steam Generator Tube Rupture Accident in a Lead-Cooled Fast Reactor
Authors: Chen Yutong , Zhang Dalin , Feng Zhenyu , Lin Yue , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: November 2024

Because of the unique design of Lead-cooled Fast Reactor (LFR), the steam generator tubes containing high-pressure water is placed directly in the high temperature lead pool, therefore once a steam generator water tube ruptures during operation, the high-pressure subcooled feed water would interact directly with hot lead, giving rise to Steam Generator Tube Rupture (SGTR) accident, which is one of the accident scenarios that must be considered during LFR design and safety analysis process. Focused on this topic, numerous experimental and theoretical research have been conducted in recent decades. In this study, the Eulerian multiphase analysis code ACENA was applied to simulate the postulated SGTR accident of European pool-type lead-cooled fast reactor ALFRED. Attention was paid to the simulation results of safety parameters, such as the evolution history of maximum cladding and fuel temperature, core relative power, core void fraction and cover gas pressure; the migration process of steam bubbles was visualized as well. Furthermore, the sensitivity of model ruptured tube quantities and neutronic feedback models were also analyzed. This study could provide reference for LFR design and safety analysis work.
Discussion on Radiation Protection for PWR Second-Loop Steam Supply Outside for Industrial Application
Authors: Yaping Zhuang , Yuan Fang , Wu Fang , Miao Zhengqiang
Published: November 2024

Nuclear power units have the ability to provide large-scale, stable industrial steam. With the rising prices of fossil fuels, nuclear energy has become economically competitive in supplying industrial steam. Laboratory measurements show that, under the condition of no leakage in the heat transfer tube of the steam generator, the concentration of radionuclides in the condensed water of the secondary loop steam is not different from that in the environmental water sample except for tritium. The IAEA technical report provides a two-cycle scheme of PWR supply steam for SAGD, which has the advantages of simple system, less process equipment, and high steam quality compared to the external steam supply scheme from the tertiary loop. Based on the drinking water standard, this article proposes to control the target value of radionuclide management for industrial application of secondary loop steam at 3.0 Bq/kg; Add a high-sensitivity radiation monitor on the steam generator blowdown line, and carry out technical renovation on the main steam pipeline N-16 radiation monitoring channel to provide timely, effective, and reliable monitoring of steam generator heat transfer tube leakage. It is interlocked with the control valve at the boundary of the external steam supply pipeline factory area to prevent radionuclides from passing through the steam supply pipeline factory area during operation and accident conditions.
Development of a New Crust Model for Analyzing VULCANO VBS-U3 MCCI Experiment With MPS Method
Authors: Yamada Takeshi , Li Xin , Yamashita Takuya , Yamaji Akifumi
Published: November 2024

In the postulated severe accident of a light water reactor, the molten core material (corium) may be released to the reactor containment floor and the Molten Core - Concrete Interaction (MCCI) may take place. This complex phenomena may involve dehydration, thermal decomposition, and ablation of concrete, redox reactions of oxidic and metallic phases of corium, solidification and remelting of corium, mixing of molten concrete with corium and so on. Evaluation of the resulting concrete floor ablation is important as it may threaten the containment boundary.

The Moving Particle Semi-implicit (MPS) method discretizes fluids and solids with calculation points (particles) in a Lagrangian manner and has been extensively developed for analyses of various MCCI tests with consideration of crust formation at the melt-concrete interface. Due to its Lagrangian nature, there is always some randomness to each particle position and Particle Shifting (PS) is one of the numerical techniques, which corrects such randomness by slightly shifting the particle position from a particle dense region to a coarse region to sustain numerical accuracy of the MPS method. Such slight correction does not usually have any physical impacts on the calculation unless the particle velocity becomes extremely small and duration of the phenomena becomes extremely long.

In this study, a new crust model is being developed to analyze MCCI, which involves continuous concrete ablation with presence of the crust layer between the corium and the concrete walls, which may gradually move with the slow concrete wall ablation process over long time. The new crust model must enable accumulation of physical displacement of the crust particle over long time (i.e., enable physical creeping) while preventing accumulation of numerical displacement of the crust particles over long time (i.e., preventing numerical creeping), Hence, in the new crust model, the PS has been effectively disabled for the crust particles. Qualitative validity of such numerical modeling was confirmed through some trial analyses of VULCANO-VBS test using a set of tentative calculation conditions and parameters, which should be carefully revised for future quantitative discussions including validation of the analysis results with experimental results.