bluuebunny/crossref_metadata_2025_split huggingface dataset:
Dynamic Event Tree Analysis of SGTR Accidents Considering HRA
Authors: Qu Nianchun , Wang He 
Published: November 2024

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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.