bluuebunny/crossref_metadata_2025 huggingface dataset:
Study on Stabilization Technology of Base Isolation Layer Using Air Floating Technology by Shaking Table Test
Authors: Takuto Aramaki , Furuya Osamu , Hiyama Yoshiro , Yamazaki Koji
Published: November 2024

Japan is one of the world’s most earthquake-prone countries, and in facilities dealing with ultra-fine processing equipment, there is a focus on maintaining system functionality during earthquakes below a certain magnitude to prevent facility damage during major earthquakes. This emphasis aims to reduce downtime and production losses by minimizing equipment stoppages and facilitating restarts after earthquakes below a certain magnitude. To meet these performance requirements, the development of new seismic isolation technologies is being sought. While air floating units exhibit high vibration isolation performance against horizontal input, concerns have been raised about the stability of load support performance in the vertical direction [1,6]. To address this, additional spring elements have been introduced to maintain the load support performance of the air floating unit. However, there are concerns that the introduction of spring action may worsen the vibration conditions of the seismic isolation target due to rocking vibrations. This paper introduces additional spring elements to suppress rocking vibrations on the upper part of the air levitation unit and presents the performance evaluation results from vibration table experiments. As a result of adjusting the additional spring, it was confirmed that rocking vibrations could be suppressed even during simultaneous vertical and horizontal input, maintaining horizontal seismic performance. Additionally, the analytical model was validated against the experimental results, confirming the validity of the analytical model.
Study on the Characteristics of Nuclear Fuel Industry
Authors: Shi Lei , Wang Yihan , Liu Hongjun , Zhang Honglin
Published: November 2024

The nuclear fuel industry is related to high-tech strategy and energy security. It is an important technical and material foundation for China’s nuclear power development. In the new era, in order to better promote the high-quality development of China’s nuclear fuel industry, it is necessary to summarize historical experience, further clarify the path of development, carefully study and deeply grasp the characteristics of the industry itself, and on this basis, provide correct guidance for the development of China’s nuclear fuel industry. This study summarizes the development experience of the nuclear fuel industry in major countries, and objectively analyzes the characteristics of the nuclear fuel industry from multiple aspects such as country, technology, enterprise, and market. Study has shown that nuclear fuel has the characteristics of a “pyramid” structure, with the top-level features of nuclear deterrence and international nuclear safeguards.
Application of Virtual Reality Technology in Human Factors Verification of Nuclear Power Plant Design
Authors: Hao Dong , Fan YanFang , Bai BingHe
Published: November 2024

This paper introduces a Virtual Reality (VR)-based Human Factors Engineering (HFE) framework for enhancing nuclear power plant designs through quantitative assessments, employing VR simulations of construction and maintenance to reduce design revisions. The initial phase entailed the creation of an integrative tool leveraging Unity and C#, which substantially mitigates the complexities associated with migrating from conventional modeling environments to VR visualizations. Subsequent stages involved the adoption of advanced virtual design methodologies and motion capture technologies to systematically capture and analyze human kinematics and interactions within simulated settings. Following this, a multi-dimensional evaluation system has been customized specifically for the nuclear environment, including operability, comfort, special environments. The fuzzy comprehensive evaluation method is used to conduct quantitative evaluation based on the established evaluation system and the data captured by the system. This methodology enables the anticipation of potential ergonomic issues at the design stage, allowing for iterative improvements and the generation of optimized solutions. Furthermore, the paper exemplifies the practical application of this method through two case studies: the selection of steel cage joints during construction and the optimization of evaporator compartment layouts during maintenance, illustrating the method’s efficacy. This comprehensive approach is designed to be applicable across the entire spectrum of design phases, from conceptualization to detailed execution, thereby addressing the distinctive challenges inherent in nuclear power plant engineering. Collectively, this research contributes to the enhancement of design quality, cost-efficiency, and operational efficacy in the field of nuclear power plant design.
Development and Optimization Criteria of Cementitious Mortars Used for the Immobilization of ILW Radioactive Waste
Authors: Rizzo Francesco , Rosa Domenico , Mangialardi Teresa , Di Palma Luca , Giannetti Fabio
Published: November 2024

The cementing process is the most widely adopted method of intermediate-level waste (ILW) conditioning, involving the use of cement mortars to waste incorporation. The main challenge in processing these mortars lies in achieving an optimal balance among their components. Accordingly, this study aims to investigate the essential criteria for developing prototype cement mortars suitable for immobilizing Intermediate Level Waste by investigating the effect of key parameters, such as water/cement and sand/cement ratios, on the workability of the compound. Different types of cement (CEM I, CEM II and CEM IV) were used, also analyzing the effect of aggregate particle size composition on workability, and the benefits of adding superplasticizer to reduce the water/cement ratio and ensure high mechanical strength were evaluated.

It was observed that increasing the water/cement ratio enhanced workability, while increasing the sand/cement ratio decreased it. The use of superplasticizer decreased the water/cement ratio with the same workability. In addition, the grain size composition like that provided by Andreasen improved workability compared with the as-is composition. The investigation and the integrated multifactorial approach allow for a rapid convergence towards the production of a prototype mortar with the desired characteristics for the immobilization of Intermediate Level Waste.
A Stabilization Strategy of Multi-Resolution Multiphase MPS Method
Authors: Zhong Yubao , Li Sijun , Chen Ronghua , Guo Kailun , Tian Wenxi
Published: November 2024

CFD simulations of multiphase behavior are essential in the safety analysis of nuclear reactors. As a particle method, the MPS method inherently possesses the capability of capturing interfaces or free surfaces. The PArticle method for Nuclear reactor thermal hyDraulic Analysis (PANDA) is a software platform based on particle methods, aiming to provide a safety analysis of key thermal-hydraulic phenomenon in nuclear reactors. The integrated PANDA-BD adopts the specially designed Multiphase MPS method (MMPS method), and shows great ability in difficult multiphase simulation.

However, due to the semi-implicit algorithm and the uniformity of particle size, MMPS incurs a relatively high computational cost. To improve the efficiency, some multi-resolution techniques have been developed for the MPS method.

Another challenge with the MPS method could be the stabilizing issue. A seriously bias particle distribution could lead to the failure of the simulation. Most of the current rearrangement methods assume a uniform particle size, and are therefore not suitable under a multi-resolution scheme.

In this study, the concept of local compressed ratio (LCR) is proposed to measure the uniformity of the particle distribution under multi-resolution. An optimized particle-shifting technique based on the novel interface recognition and LCR is developed. A series of regularization tests are performed, showing the regularization ability under both single and multiple resolution cases. Simulations of Rayleigh-Taylor Instability are presented, the novel interface detection is tested and shows good accuracy, the robustness and efficiency of the proposed method for multiphase simulation is proven.
Application of Global Importance Measures in RISMC for Dynamic Risk Assessment
Authors: Ma Liye , Wang He , Wang Longcong
Published: November 2024

Risk-Informed Safety Margin Characterization (RISMC) integrates traditional deterministic safety analysis methods with probabilistic safety analysis approaches by employing stochastic sampling of uncertain parameters and dynamic branching simulations of nuclear power plants. It enables time-dependent, uncertainty-aware, and multi-level risk assessments. In the process of RISMC analysis, quantifying the impacts of uncertainty parameters, response events, and system components on nuclear power plant risk can support the implementation of risk analysis results in nuclear power plants. However, traditional importance measurement methods cannot directly apply to RISMC. This study adopts the global importance measurement method to measure the time-dependent importance. The selected global importance measurement method considers input parameter uncertainties and uncertainties from interactions between input parameters. It is demonstrated on Station Blackout (SBO) accidents modeled with dynamic event trees and compared with traditional importance metrics used in probabilistic risk assessment. The risk importance of systems and physical design parameters in SBO accidents is ranked. Results indicate timely AC power recovery significantly reduces risk, while the importance of TAFW failure diminishes over time. Future research is needed to investigate the discrepancies between the global importance measurement method and traditional methods.
Modelling Mass Transfer and Mixing of Corium with Concrete during Molten Corium-Concrete Interaction: Impact of Concrete Thermal Conductivity
Authors: Khurshid Ilyas , Addad Yacine , Afgan Imran
Published: November 2024

Molten corium, a mixture of nuclear fuel, claddings, and structural elements, would form after the meltdown of a reactor core in the event of a severe accident at a nuclear power plant. This paper presents the development and implementation of a new numerical mixing model in OpenFOAM for a comprehensive analysis of concrete ablation during the Molten Corium-Concrete Interaction (MCCI) process. The well-researched CCI-2 experiment served as the model’s benchmark during model validation. The model can forecast heat transfer, concrete ablation, phase change, mass transfer, and mixing of concrete and corium. Additionally, the effect of concrete thermal conductivity was investigated to study its impact on concrete ablation. We observed significant low concrete ablation, controlled temperature, velocity and density field for concrete with low thermal conductivity. Accordingly, the ablation of concrete decreased by 61 and 58 %, when the thermal conductivity was decreased to 1.5 and 1.0 W/m.K, respectively from 2.0 W/m.K. Our results also revealed that the melted concrete and corium, even though, having different densities, form a single phase/mixer that could enhances the ablation of concrete.
Dynamic Event Tree Analysis of SGTR Accidents Considering HRA
Authors: Qu Nianchun , Wang He
Published: November 2024

The development of nuclear power is of significant importance in meeting the growing energy demands. The ongoing trend With a paramount Emphasis if safety is evident. Steam generator tube rupture (SGTR) accidents, characterized by a relatively high frequency and demanding operational requirements for operators, often involve simplified modeling of human factors in dynamic event tree (DET) method. This approach may struggle to depict the impact of operator error probability on DET sequence probabilities under different scenarios. A suitable human reliability analysis (HRA) is crucial for a more accurate assessment of system safety. This paper uses the DET method and standardized plant analysis risk-human reliability analysis (SPAR-H) method to analyze the SGTR accident of the CPR1000 unit, by calculating the operator’s maximum allowable time in different DET sequence scenarios, and studies the impact of extending the maximum allowable time of operator operation on human error probability (HEP) under different DET sequence scenarios, and addresses the limitations of simplified operator modeling in DET method. Analyzing DET sequence probabilities from a more realistic perspective demonstrates the advantages of incorporating SPAR-H for human reliability analysis within the DET method. The work is valuable for nuclear safety assessment.
Thermal-Hydraulics and Neutronics Coupling Calculation and Validation of NECP-Panda: A Computational Code for Pebble-Bed High Temperature Gas-Cooled Reactors
Authors: Xu Dongyu , Wang Yongping , Wu Hongchun , Zhang Aolin , Wu Yuxuan , Luo Yong
Published: November 2024

In recent years, the Pebble-bed High Temperature Gas Cooled Reactor (PB-HTGR) has become a hot area in the research of the fourth-generation nuclear power system. The calculation of temperature distribution in the core and fuel sphere is one of the key tasks in the thermal design and safety analysis of the core. Due to the cylindrical coordinate characteristics of HTGR and the uncertainty of pebble flow in core, the complexity of coolant flow paths and the need to cover both the core and reflector regions in thermal calculations, accurate simulation of thermal hydraulics poses challenges. The mainstream PB-HTGR design and analysis codes use the fine-mesh finite difference method to solve the core physics and thermal hydraulic problems, and only has the ability of two-dimensional calculation and analysis. In order to address challenges in the core design and optimization of PB-HTGR and improve the accuracy of simulating the core state of the PB-HTGR during high temperature operation, the Nuclear Engineering Computational Physics (NECP) laboratory of Xi ’an Jiaotong University has independently developed NECP-Panda, a physical and thermal hydraulics calculation code specially designed for PB-HTGR. In the thermal hydraulic calculation of the core, NECP-Panda solves the solid heat conduction equation by using the coarse mesh nodal expansion method (NEM) to obtain the three-dimensional temperature distribution of the solid. The fine-mesh finite difference method is used to solve the convection diffusion equation, and the three-dimensional pressure field distribution and the three-dimensional mass flow field distribution of gas are obtained. To ensure convergence, NECP-Panda incorporates the exponential approach method to handle gas energy conservation, and employs an assumption of linear solid temperature distribution between nodes to solve the gas temperature distribution in the core and flow channels. The three-dimensional temperature field and three-dimensional pressure field are iteratively solved using the overrelaxation method until convergence. The cross section of the physical calculation is updated according to the three-dimensional temperature distribution to obtain more accurate calculation results. To validate the computational capability and accuracy of NECP-Panda for nuclear-thermal coupling, core power and temperature distribution were computed using the High Temperature Reactor Pebble-bed Module (HTR-PM) model of a HTGR nuclear power plant demonstration project. The power and temperature distributions obtained from NECP-Panda exhibit good agreement with those calculated by VSOP, a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. Therefore, NECP-Panda has good computational power and accuracy for the thermal hydraulics calculation of PB-HTGR.
Analysis of Operational Characteristics of a Small Modular Reactor With Accident Tolerant Fuel
Authors: Lachout Ondřej , Suk Pavel
Published: November 2024

The topic of the paper is focused on numerical simulations of nuclear reactors, the preparation of macroscopic nuclear data using a deterministic approach, and the analysis (with an optimisation proposal) of selected Accident Tolerant Fuel (ATF) concepts of fuel pellets. The behaviour of ATFs in large cores has already been investigated in depth, but analyses in small cores have not yet been studied in more detail. Macroscopic nuclear data were prepared using the SCALE-Triton computational code on an infinite fuel assembly model. Subsequently, a 3D full-core calculation is performed in the deterministic macrocode PARCS on the model of a light-water Small Modular Reactor (SMR) developed by the NuScale Power company. For proper interpretation of results, coupling of PARCS with an external thermohydraulic solver PATHS, is studied. The behaviour of selected ATFs and deviations of various neutronic parameters from the referenced UO2 fuel are analysed in this complex model. Together with the analyses, a proposal on the optimisation of the referenced core with the (U10Mo/UN/ThO2) pellets combination for the use on this model, is presented. This optimisation proposal brought the possibility of extending the fuel campaign (while maintaining the same average fuel enrichment), reducing the average fuel temperature and improving the inherent safety parameters, at the expense of a slight increase in the value of the maximum peaking factors.
TRISO Particle Behavior Analyses of an HTR-10 Type Reactor Employing the OTTO Operation Scheme
Authors: Zhang Junxiao , Liu Songyang , Xie Junhao , Zhang Tian
Published: November 2024

High-temperature gas-cooled reactors (HTGRs), renowned for their safety features and capacity to provide high-temperature heat, are anticipated to find widespread applications. The HTR-10 serves to demonstrate the intrinsic safety characteristics of modular HTGRs. Different refueling schemes result in varied pebble bed flows, with the Once-Through Then Out (OTTO) scheme causing fuel pebbles to remain in the core for longer periods. This increased residency in the core exposes the pebbles to more intense operating conditions, especially those pebbles that traverse the core’s center. In this work, the VSOP code coupled with THERMIX is used to calculate the pebble bed power distribution and radius-wised graphite temperature while tracing the flow of representative pebble and TRISO. Obtained data are fed to a 1-D OFFBEAT TRISO model to simulate the temperature distribution and variation, as well as the deformation of each layer of the particle. By this method, the fuel temperature, mechanistic properties, and fission gas release during the OTTO cycle can be estimated. The results show that in the core center, the fuel temperature reaches over 1100K while at the edge of the core is about 1050K, well below the criteria of 1473K. Mechanical analyses show that the maximum equivalent stress of TRISO particles in the core is about 350 MPa with an FGR of about 17.5%. This work offers a foundation for further analysis of fuel failure probability and fission product diffusion and release under normal operating conditions. This work established a method to analyze the fuel behavior of HTR TRISO fuel using legacy code and high-fidelity solver and play as a foundation for further fuel failure and fission products diffusion analyses.
NPP Modification Opportunity by Hydrogen Technologies Implementation: Assessment in Frame of the Euratom Project NPHyCo
Authors: Kykal Martin , Mazurok Oleksandr , Glueckler Martin , Herrero Moriana Cecilia , Serin Canet , Korolchuk Vasyl
Published: November 2024

NPHyCo (Nuclear Power and Hydrogen Cogeneration) is an EU research project dedicated to the production of hydrogen from nuclear power. The project is funded by the EU’s Euratom Research & Training program (2021–2025) dedicated to nuclear research and innovation. NPHyCo started in the Autumn of 2022 and will last for two and a half years.

The EU hopes to fully decarbonize its economy by 2050 [1]. One of the solutions being put forward is hydrogen. Nevertheless, to date most of the hydrogen produced in Europe comes from fossil fuels [2]. So the question is how to increase the production of low-carbon hydrogen to ensure that enough is being produced — at an affordable cost — to meet the needs of hard to decarbonize sectors, such as industry?

The article presents the structure of the project and the results achieved so far by the different working groups of the NPHyCo project, such as concept, technology, economics, licensing, implementation, selection of a suitable location for the hydrogen production plant (HPP) and communication.

Particular attention is paid to the nuclear power plant (NPP) modification in the case of a hydrogen facility construction, its connection to the NPP resources and NPP safety assessment.

The first results regarding Ukrainian NPPs were presented in [3], but since then new results have been obtained, which are presented in the next sections. Since the largest scope of information was collected for Ukrainian NPPs we use the Rivne NPP (RNPP) for the best tuning of all universal models in WP3, WP4 and WP5. Therefore, the results for this NPP in particular are presented below.
A Modified Pseudopotential Lattice Boltzmann Model With Improved Thermodynamic Consistency for Circular Interface
Authors: Li Shunyang , Wan Li , Gui Nan , Yang Xingtuan , Tu Jiyuan , Jiang Shengyao
Published: November 2024

The pseudopotential model offers a straightforward solution for multiphase flow. However, it falls short in precisely predicting thermodynamic density and pressure for droplets and bubbles, potentially introducing errors in multiphase flow simulations. This paper introduces a modified pseudopotential lattice Boltzmann method to address the limitation. The model incorporates the equation of state by introducing an external force and modifies the effective mass to ensure thermodynamic consistency, eliminating the need for additional parameters to adjust the equilibrium density of liquid and vapor phases. The model’s performance is validated through benchmark tests and compared with other models. The accurate calculation of the equilibrium density of liquid and vapor for flat interface demonstrates that the model satisfies the thermodynamic consistency. The consistency between surface tension calculated by Young-Laplace law and by an integration across a flat interface is confirmed. The model produces the expected equilibrium thermodynamic pressure for both droplets and bubbles. In addition, the model suppresses spurious velocity by confining its region to near the interface. The features render the model capable in simulating multiphase flow, particularly for scenarios where thermodynamic pressure plays a crucial role.
Study of Compressible Flow in Pipes Containing Mesh Structures
Authors: Xu Jun , Yu Hongxing , Deng Jian , Liu Yu , Zhang Luteng , Tian Langlang , Zhong Lei
Published: November 2024

In the study of compressible vapor flow in pipes with mesh structure, there are few papers about the effect of mesh structure on vapor friction, especially the empirical formula of friction coefficient is lacking. In this paper, a vapor flow experiment in a pipe with the mesh structure is designed, and the friction effect of mesh on the vapor in the pipe is studied. According to the experimental results, the friction coefficient relationship is fitted, and the compressible vapor flow model is constructed based on the friction coefficient relationship. The results show that the friction resistance of the mesh structure to the vapor in the pipe can not be ignored, and the greater the vapor flow rate is, the more obvious the fricton effect of the mesh is.
Microscopic Turbulence Topology Analysis of the Cross Flow Over a Square Arranged Tube Bundle Using Direct Numerical Simulation
Authors: Xie Xiaoyang , Zhao Houjian , Li Xiaowei , Wu Xinxin
Published: November 2024

Helical tube bundles with small tube pitches are usually preferable for intermediate heat exchangers (IHXs) of High Temperature Gas-cooled Reactors (HTGRs) for its compactness. Too small tube pitches would introduce more complicated flow features in cross flow over tube bundles. In the current investigation, microscopic analysis on turbulence topology of cross flow over a square arranged tube bundle are investigated using Direct Numerical Simulation (DNS). The time averaged flow field as well as the Reynolds stress distributions are first discussed. Then, a method that utilizes a red-green-blue (RGB) color map to depict the anisotropic states of Reynolds stress are used to analyze the overview Reynolds stress state of cross flow over tube bundles. Traditional Lumley triangle is also used to further investigate detailed Reynolds stress state variation in the near wall region with different pressure gradient and the shear layer. In the region with favorable pressure gradient (FPG), the Reynolds stress state changes from a two-component state near the wall towards a disk-like state (state with two equal components), and eventually reaches an isotropic state in the fully turbulent region. In the region with adverse pressure gradient (APG), the Reynolds stress state changes from a two-component state near the wall towards a rod-like state (one component), and eventually reaches an isotropic state. The Reynolds stress state also appears strong anisotropy (rod-like state) in the shear layer. However, according to one-dimensional energy spectra, the anisotropy near the wall occurs at all wavenumber ranges (all scale ranges), while the anisotropy at the shear layer only occurs at the low wavenumber range (large-scale range).
Numerical Simulation Study of Flow and Heat Transfer Characteristics of Asymmetric Microchannel Intermediate Heat Exchanger
Authors: Chu Junhao , Xu Zhe , Zhao Yanxin , Wang Xiding , Wang Wei , Liu Xinhe , Zeng Dong , Fu Wen , Li Peiyue
Published: November 2024

The combination of a lead-bismuth fast reactor and a supercritical CO2 Brayton cycle is one of the most ideal and efficient power cycles. The printed circuit heat exchanger is best suited for the system’s intermediate heat exchanger. Its thermal and hydraulic performance have a significant impact on the overall system’s circulation efficiency. There are substantial differences between the physical properties of hot and cold fluids. The traditional symmetrical flow channel structure does not adequately meet the requirements. In this paper, a microchannel intermediate heat exchanger with asymmetrical channels on both sides is numerically simulated. The study analyzed the effect of liquid lead-bismuth eutectic (LBE) inlet Reynolds number, flow channel cross-sectional shape, supercritical carbon dioxide (S-CO2) inlet Reynolds number, and inlet temperature on the thermal and hydraulic performance of the intermediate heat exchanger (IHX). The study investigated the effect of different parameters on the Nusselt number, heat transfer coefficient, and Fanning friction factor on both sides of the heat exchanger. The performance of the intermediate heat exchanger was assessed using performance evaluation criteria. The results show that the IHX performance is insensitive to changes in the Reynolds number and cross-section shape of the hot side flow channel. The heat transfer performance of the cold side is significantly inferior to that of the hot side. Changing the cold side parameters can affect the performance of the IHX to a greater extent. Increasing the inlet temperature of the S-CO2 has less impact on the performance of the IHX than increasing the Reynolds number.
The Neutronics Analysis of Heat Pipe Cooled Traveling Wave Reactor Designs With High-Assay Low-Enrichment Uranium and Natural Uranium
Authors: Hu Po , Ma Kunfeng
Published: November 2024

The heat pipe cooled traveling wave reactor (HPTWR) is a novel design. This paper firstly presented a HPTWR design featured with a traditional traveling wave reactor core, which consists of high-assay low-enrichment uranium (HALEU) section (20%) and natural uranium section (0.725%), or ignitor section and breeder section. The core was a hexagonal prism and assembled with more than a thousand of separated UN fuel elements. Each fuel element was a smaller hexagonal prism with a heat pipe channel in the center. A lithium and Mo/Re heat pipe was used in the current design. The neutronics calculations were performed using Monte Carlo code RMC with heat pipes, cladding and UN fuel had a same operational temperature (1700 K). And the results showed that at the beginning of the cycle reactivity is high, and Keff has reached 1.36, and power is concentrated in the ignitor section with nearly zero power in the far end of breeding section. Then an improved HPTWR design is proposed to limit the swing of keff less than 1%, and distribute the power more evenly by reducing the enrichment difference between ignitor section and breeder section with similar core geometry. The HPTWR using natural uranium can reach 100MWth of power and 350 years of refueling cycle, and the improved HPTWR can reach 70MWth and more than 40 years of refueling cycle.
Multi-Phase Flow Dynamic Analysis Under Large Leakage Sodium-Water Reaction Accident in a Sodium-Cooled Fast Reactor
Authors: Bai Xi , Sun Peiwei , Wei Xinyu
Published: November 2024

The sodium-cooled fast reactor (SFR) is a promising generation-IV design with the nuclear system using liquid sodium as a coolant, high uranium utilization, and a closed fuel cycle. When the double-end-guillotine (DEG) heater tube broke in the steam generator of SFR, a high-pressure and high-temperature hydrogen bubble grew, with a large leakage sodium-water reaction (SWR) occurring. The reaction not only caused the pressure and temperature to increase promptly but also diffused the multi-phase flow (sodium-hydrogen-sodium-sodium hydroxide) into the steam generator and the secondary loop, which threatened the integrity of the secondary loop and the safety of the key equipment, such as the steam generator, surge tank, and protection system.

In this study, a Multi-phase hydrodynamic Analysis code for laRge Leak sodIum-water reactioN (MARLIN) calculating one-dimensional compressible multicomponent multi-phase flow with sodium-water chemical reaction was developed. The continuity equation, momentum equation and energy equation were applied to derive the MARLIN model. The hyperbolic-type partial differential equation was solved by the method of characteristic. FORTRAN language was used to develop the code. Critical equipment in the secondary loop model includes the steam generator, surge tank, pump, rupture disks, and primary and secondary discharge accident tanks. The ACENA code results calculating the sodium-air flow were used to verify the multi-phase code. A good agreement was reached between the ACENA code and the MARLIN code.

Based on the structure of the sodium-water reaction experiment, the developed model was applied. The tendency of calculated pressure is consistent with the experiment date. The comparison of the results between the MARLIN model and experiment data was performed to indicate the higher accuracy of multi-phase flow dynamic analysis. Meanwhile, the comparison of the calculation between the MARLIN code and the conservation sing-phase code was proposed. It can be found that the multi-phase sodium-hydrogen-sodium hydroxide flow result was more reasonable and accurate in simulating and analyzing the actual large leakage SWR dynamics in the SFR.

As a preliminary best estimate code in this study, the MARLIN code is valuable for steam generator design and protection from the large leakage SWR accident in the SFR.
Comparison of EQL0D V2 and V4 Procedures by Means of Equilibrium U-Pu and Th-U Cycles in 16 Selected Reactors
Authors: Křepel Jiří , Borys Francis , Sabol Josef , Losa Evžen
Published: November 2024

Three papers related to the equilibrium fuel cycle in 16 selected reactors were submitted to ICONE 31 conference. In this first paper, the major equilibrium cycle characteristic obtained by EQL0D v4 are compared to the previous results simulated by EQL0D v2. These older simulations were accomplished for infinite lattice and neglected fission products. Therefore, the results presented in this paper apply the same simplification to enable a cross-comparison. The impact of leakage from finite core and of the fission products are addressed in the second and third paper from this series.

The same 16 reactors are simulated, 8 with thermal and 8 with fast neutron spectrum as in the existing study. EQL0D v2 simulations use only the ENDF/B-VII.0 library, while EQL0D v4 can choose between eight different libraries. To allow cross-comparison, results are presented for the ENDF/B-VII.0 cross-section library. To demonstrate the new capabilities, results are also shown for the ENDF/B-VIII.0 library.

Even though, the EQL0D v2 and v4 substantially differ in the treatment of cross-sections provided by the Serpent 2 code and rely on different version of this code, their results are very similar. The excess reactivity, its nuclide-wise break-down and other parameters like migration area are in good agreement.

The difference between EQ0D v4 results for ENDF/B-VII.0 and ENDF/B-VIII.0 cross-section libraries is larger. This disparity is mainly caused by the lower 233U capture cross-section in the newer library, which results in decreased build-up of higher actinides nuclides.
Experimental Investigation on the Heat Transfer Capacity of a Prototypically Long Straight Thermosyphon Bundle for Nuclear Spent Fuel Pool Passive Cooling
Authors: Cáceres Castro Sergio Iván , Kulenovic Rudi , Starflinger Jörg
Published: November 2024

The Fukushima Daiichi Nuclear Power Plant disaster in 2011 underlined the need for passive residual heat removal systems in Nuclear Spent Fuel Pools (NSFPs). Two-phase closed thermosyphons (TPCTs) have been identified as a viable solution for enhancing safety in nuclear power plants by providing passive cooling during emergencies. This paper reports on the heat transfer behaviour of a 10-meter-long TPCT pipe bundle, tested at the ATHOS (Atmospheric THermosyphon cOoling System) facility, University of Stuttgart. The bundle, arranged in a 3 × 3 quadratic format with vertical tubes, includes a 2-meter evaporator and a 4-meter condenser section. Based on preliminary findings from individual tests on a single pipe TPCT in the THOR (THermosyphon labORatory) test stand which identified a 30% optimal filling ratio for the 2-meter evaporator section, experiments were conducted under realistic NSFP temperatures (45°C, 60°C, 80°C) and heat-sink temperatures (5°C to 30°C) with an average airflow velocity of 0.4 m/s. The results demonstrated the TPCT bundle’s ability to dissipate up to 6.5 kW at 80°C, affirming its efficiency in passive NSFP cooling. Start-up tests, progressively heating the water pool from 30°C to 80°C, validated the reliability of the system and the effectiveness of improvements such as optimal filling ratio, TPCT degassing, and enlargement of the evaporator section, which collectively enhanced the heat transfer rate by approximately 30% compared to prior studies.
Flow Dynamics in a Fuel Bundle With a Particle Bed
Authors: Franz Scott C. , Wissinger Gordon J.
Published: November 2024

The pressure reduction in the RCS and temperature excursion of the cladding during a postulated loss-of-coolant accident (LOCA) can lead to cladding rupture. Fuel fragmentation can occur at any burnup, but the fuel fragment sizes decrease with increasing burnup, thereby increasing the potential for relocation and dispersal via the rupture opening. Research is ongoing to better understand fuel fragmentation, relocation, and dispersal (FFRD). However, the research to date is focused largely on understanding the mechanisms generating the fragmentation and quantifying the resulting fragment sizes and amounts within the fuel rod. There is a need to understand the consequences of the dispersal from the fuel rods. Some of the dispersed particles can settle onto a spacer grid below the rupture location forming a particle bed. Framatome is researching the behavior of fluid flow through and around a particle bed that has the characteristics of fuel particles that have been dispersed from rupture locations within a fuel bundle. Understanding the behaviors and changes in the flow patterns caused by the presence of the particle bed are of importance as they relate to the cooling behavior of the fuel pins during a LOCA transient. The presence of the particle bed can either divert steam and droplet flow away from the location near the particle bed region or may slow the flow of steam and droplets through the particle bed. A reduction of flow in the local region of the particle bed may adversely impact the cladding temperature during the transient. Tests for the benchmarking of flow patterns and cooling dynamics for these kinds of conditions are limited. Consequently, Framatome has performed full scale fuel bundle flow testing to provide needed benchmarking data. The testing includes bundle geometrical aspects under scaled LOCA conditions which include clad swelling, rupture locations, spacer grids, mixing grids, the presence of a particle bed, and entrance and exit effects. Multiple sizes of particles including mixtures of sizes are taken into account as well as multiple bed depths and different bed formations in the testing. Additional testing has been performed to establish baseline characteristics of fuel particle beds in general (fluidization behavior, packing fractions, etc.). The data collected has provided results leading to in-depth understanding of the flow patterns and pressure drops in and around the particle bed with a high level of detail. This data can then be used to develop and benchmark tools capable of simulating the flow patterns within a nuclear fuel bundle with a particle bed. An increased understanding of how steam and droplets interact with a particle bed in the geometrical conditions found in a nuclear fuel bundle is a requirement for developing new safety methods.
Manufacturing Challenges for Hot Isostatic Pressing Large-Scale Vessels
Authors: Sulley John , Stewart David
Published: November 2024

Hot Isostatic Pressing (HIPing) has been used by Rolls-Royce to successfully manufacture nuclear plant components such as: valves, piping, and pump casings, the majority of these components being manufactured from stainless steels, typically 316L. There are potential benefits to be gained by manufacturing large nuclear plant pressure vessels via the HIP process, such vessels commonly being manufactured from Low Alloy Steel (LAS) materials such as ASME SA-508. HIPing of large-scale LAS structures present specific challenges. This paper presents the challenges Rolls-Royce has experienced in their HIP vessel development work and means of addressing those challenges. It covers key steps of the HIP manufacturing process, from powder production, through can manufacture including leak detection and HIP consolidation. Of particular note are the risks of powder oxidisation and contamination, achieving the necessary material toughness, and can failure during the HIP cycle as a consequence of poor can design and fabrication. The paper shows examples of the results of failures, e.g. contaminated and unconsolidated powder, and how the risk of such failures can be reduced.
Thermal Performance Analysis of Radioactive Material Transport Containers
Authors: Wang Changwu , Zhang Yuhang , Lian Yiren , Chen Lei , Wang Zhipeng , Sun Qian , Zhuang Dajie
Published: November 2024

The safe transportation of radioactive materials is a critical component of nuclear safety, with transport containers acting as the primary safety barrier during these operations. Thermal performance analysis plays a critical role in evaluating the safety performance of these containers. In our study, we conducted thermal tests and simulation analysis on AF transport containers. The focus was primarily on comparing the simulated maximum temperature results of key components with the actual test outcomes, after simulating transport accident conditions. The comparison revealed a strong correlation between the simulation analysis results and the experimental data, with an error margin of less than 14.4%. These findings demonstrate that the combined application of simulation analysis and experimental methods for collecting temperature data of critical components within transport containers can effectively evaluate the thermal safety performance of radioactive material transport containers, thereby enhancing nuclear safety standards.
Effect of the Heat Transfer Correlation Choice on the Coupled Assessment of Energy Release and Stress-Strain State of an Elementary Supercritical Fuel Rod Under Nuclear Heating Conditions
Authors: Filonov Vladislav , Kovalenko Oleksandr , Dubyk Yaroslav , Filonova Yuliia , Fedorov Dmitriy
Published: November 2024

The paper presents a multiphysics procedure for estimating the parameters of a fuel channel under nuclear heating conditions. The idea of the approach is to determine the multiplying properties of an elementary fuel rod in the 4-group diffusion approximation, together with the estimation of the temperature field of both the supercritical coolant and the structural elements of the fuel rod. Group constants are prepared on the basis of a precision code that takes into account the design features of the reference fuel assembly. Several dozen different correlations are implemented into the procedure to calculate heat transfer. A concept of parametric energy release functions is proposed, which can be useful, for example, when modeling the core using CFD. The concept is based on the analysis of the influence of parametric temperature profiles (according to the classification of heat transfer regimes) on the effective multiplication factor. The paper presents the results of assessments of the fuel rods temperature state and thermal power depending on the selected reactivity margin and the selected HTC correlation. The stress-strain state of fuel rods was analyzed for each calculation option. The developed approaches are fast and can be used to optimize design decisions.
Analysis of the Effect of Random Bowing of the Fuel Pin on the Power Distribution in the Fuel Assembly
Authors: Závorka Jiří , Lovecký Martin , Škoda Radek , Monhartová Petra
Published: November 2024

One of the main approaches to ensuring nuclear safety in neutronic calculations is the consideration of engineering factors. These factors define the margin of the individual parameters of the power distribution in the reactor core up to their safety limit, which must be assumed in the design of the fuel loading and which is controlled during the operation of the nuclear reactor.

The resulting engineering factor consists of many components, and one component of this set is the subject of investigation in this paper, namely the evaluation of the fuel pin bow on the power distribution.

This study provides a fundamental investigation of the effect of the fuel pin bow on the power distribution in the fuel assembly without knowing the maximum bow specified by the fuel supplier or determined by measurement.

The methodological approach is based on randomly generated fuel pin deflections that simulate the fuel pin deflection/bow using a specially developed program, including a few conservative assumptions.
Development of an Onset of Nucleate Boiling Model and a Nucleate Boiling Heat Transfer Model Within the Fouling Layer
Authors: Wang Xiaowen , Liu Maolong , Cong Tenlong , Gu Hanyang
Published: November 2024

The fouling on the reactor fuel cladding surface affects the boiling heat transfer, and the boiling within the fouling involves both nucleate boiling and film boiling regimes. It is crucial to establish the boiling onset model to determine the boiling regime within the fouling layer. By simplifying the fouling layer as the porous media with steam chimneys, it is assumed that the cladding temperature is exactly the saturation temperature at the onset of nucleate boiling, and the energy still relies on convective heat transfer when nucleate boiling occurs. A new model is proposed to determine the onset of the nucleate boiling within the fouling. And the wall temperature at nucleate boiling is obtained by simple linear interpolation. The errors of the onset of the nucleate boiling and nucleate boiling model from the experimental results is within 15% and 20%, respectively.
High-Order Harmonics Calculation for the Power Reconstruction of 3D Pin-Scale Whole-Core
Authors: Xie Zhinan , Hao Chen , Yin Wen
Published: November 2024

The high-order harmonics of neutron diffusion equation is used in the reconstruction method to reconstruct the core power distribution. There are two methods available for the high-order harmonics solving: power iteration method with source correction (PIMSC) and Krylov subspace method. PIMSC was commonly used in the high-order harmonics calculation of the homogeneous assembly-scale neutron diffusion equation. However, this method cannot use the discontinuity factor theory, and the diffusion system is not equivalent to the transport system. Meanwhile, the convergence rate is influenced by the dominance ratio. In order to calculate the high-order harmonics of the 3D multigroup pin-scale whole-core neutron diffusion equation and guarantee equivalence between the neutron diffusion equation and the neutron transport equation, the 3D pin-scale generalized equivalence theory based coarse mesh finite difference (gCMFD) system is used in the high-order harmonic calculation, which is equivalent to the transport system using the discontinuity factor theory. The implicitly restart Arnoldi method (IRAM) is used to solve the 3D pin-scale gCMFD system. IRAM is a kind of Krylov subspace method which utilizes the implicit restart technology to improve the solving speed. To verify IRAM in high-order harmonics calculation, the results of C5G7 benchmark calculated by the PIMSC and IRAM are compared. The whole-core model of low-temperature heating reactor (LTHR) is used to test the calculating ability of pin-scale high-order harmonics and analyze the express ability of power. The numerical results indicate that IRAM has superior computational efficiency compared to PIMSC, being approximately 431 times faster. Additionally, for the low temperature heating reactor, the high-order harmonics solved by IRAM have good accuracy in the pin-scale power reconstruction.
Analysis of Shielding Structures of Radioactive Material Transport Package Based on Isotope Γ Radiation Properties in Radioactive Materials
Authors: Wang Zhi-peng , Wang Chang-wu , Sun Qian , Zhang Yu-hang
Published: November 2024

The transportation package is an important equipment for the safe transportation of radioactive materials. Many countries and IAEA have restricted the shielding performance of the transportation container of radioactive materials. The radiation shielding performance provided by different materials, and different structural designs is also an important research area in the field of radiation protection. Uranium and its isotopes are taken as analysis objects, and the overall radiation characteristics (γ spectra) of the contents are obtained by calculating method based on the γ spectra of different nuclides provided by EDNF database. This paper forms a design method for the shielding structure design process of packages, and provides reference and guidance for shielding design of radioactive goods transport cargo package.