bluuebunny/crossref_metadata_2025_split huggingface dataset:
The Investigation of Heat Transfer Deterioration in Supercritical Water Heating Systems
Authors: Chen Yu-Sen , Lee Jin-Der , Chen Shao-Wen 
Published: November 2024

Abstract
The object of this study is to investigate the heat transfer deterioration (HTD) in the supercritical water (SCW) heating systems. The present study develops a new empirical correlation for predicting the onset heat flux of HTD in SCW and are compared with the existing ones in literature. The predicted values are adopted as the basis for judging the occurrence of HTD and verified against 108 independent operating conditions of SCW collected from the literature. The results reveal that the judgment accuracy can reach 95.37%. This study also employs the ratio (RHTD) of heat transfer coefficient to a reference value, based on different methods in literature, as an indicator for distinguishing HTD and non-HTD. Through the analyses among 312 points of HTD and 135 points of non-HTD extracted from the experimental results of SCW in literature, the results illustrate that the judgment accuracy by the threshold of RHTD = 1 can be over 85%. Furthermore, this study introduces the new definition of RHTD by using the film temperature as the reference state. Through the analyses of the same data points for SCW, the setting of RHTD = 1 based on the film temperature may be appropriate as the classification criteria between HTD and non-HTD.
Thermal Analysis and Structural Design of the Main Steam Containment Penetration for Multi-Modular High-Temperature Gas-Cooled Reactor Power Plant
Authors: Zhu Chaoyi , Zhang Yiyang , Yan Huijie , Fu Jiyang , Huang Mei 
Published: November 2024

Abstract
Based on the high-temperature gas-cooled reactor pebble-bed module demonstration plant (HTR-PM), which employed two small modular reactor modules (SMRs) with thermal power of 250 MWth and power generation of approximately 100 MWe per module, several larger versions of the multi-modular high-temperature gas-cooled reactor power plant (MHTGR) with more modules are under design. The newly designed MHTGR contains six to ten modules in a large circular ventilated low-pressure containment (VLPC) for compact design. With this layout, all of the main steam pipes should pass the containment in parallel through a limited area of the containment wall. The multiple main steam pipe containment penetration design for the MHTGR should meet several essential challenges to keep the concrete wall temperature below the temperature limit. Firstly, the main steam temperature for the MHTGR is over 540 °C, one of the highest among all the commercial nuclear plants. Secondly, the wall surface area between pipes is comparatively limited in comparison to HTR-PM and other nuclear plants. The heat rejection by wall surface is restricted. Thirdly, there is no active cooling on the surface or inside the wall.
This study first set up a 2-D heat transfer model to discover the influence of the sleeve diameter and insulation thickness on wall temperature. Secondly, 3-D models were established to compute and compare the temperature distribution of potential structural design. Thirdly, the fundamental mechanical parameters were designed considering the temperature distribution and mechanical properties of the materials.
Analysis of Heat Transfer at Supercritical Conditions With Explainable Machine Learning Model
Authors: Song Meiqi , Li Haozhe , Liu Xiaojing 
Published: November 2024

Abstract
Explainable machine learning algorithm will be applied to provide accurate heat transfer prediction and information mining. Dimensionality analysis will be carried out at first in order to generate reasonable input features (dimensionless parameters) to the neural network. The supercritical heat transfer dataset with water as coolant is used for neural network training, in which 80% of the data set is randomly partitioned as the training set, while the remaining 20% constitutes the test set. Z-Score standardization is applied to preprocess the data on both the training and test sets. Following this, the neural network will be built to predict the Nusselt number with the particle swarm optimization algorithm using the supercritical heat transfer dataset and selected dimensionless parameters. Finally, the mechanism of heat transfer deterioration will be analyzed by applying the explainable algorithm for global and local model explanations. The study investigates the phenomenon of supercritical heat transfer excitation based on the changes in feature importance under different operating conditions.
Verification of the HPR1000 Safety Classification in a Probabilistic Way
Authors: Yang Benlin , Zhu Rongya , Wang Tianyue , Zhao Guanghui , Zheng Junming , Hu Lingsheng , Gong Yihong 
Published: November 2024

Abstract
Probabilistic Safety Assessment (PSA) insights according to nuclear power plant (NPP) risk importance distribution profiles are used as a complementary way of deterministic safety analysis in verifying the adequacy of the safety classification. Fractional Contribution (FC) importance measure and Risk Achievement Worth (RAW) importance measure are chosen as two key indexes to show the full scope of NPPs risk profiles. It is expected that the equipment ranked with high safety class would be in the risk importance distribution areas with high FC importance or high RAW importance, while the equipment ranked with low safety class would be in the risk importance distribution areas with low FC importance and low RAW importance. The equipment whose safety classification do not match with the risk importance distribution are identified for further assessment to understand the reasons for it. Some case studies are performed on HPR1000 to show the process of this safety classification verification method, and to show whether the HPR1000 safety classification match with its PSA results or not. The result shows that the majority of the HPR1000 equipment safety classification matches with its risk importance distribution, which provides the confidence that the HPR1000 safety classification is mainly correct and adequate. There are still some exceptions, so two typical systems are selected for further discussion to provide detail insights about the reasons for it, and some recommendations for the safety classification are then given.
Establish and Analysis of an Optimized System for Hydrogen Production From Nuclear Energy in China
Authors: Ping Wang , Fan Li , Luo Yibo 
Published: November 2024

Abstract
Addressing climate change poses a severe challenge to the energy development of countries around the world. In order to realize the Paris Agreement, energy transition has become an important task for China. Hydrogen and nuclear energy is important part of China’s new energy system in the future.
Based on the dilemma of hydrogen energy development in China, an optimized hydrogen production system suitable for the coupling development with China’s nuclear energy is proposed.
The result shows that the hydrogen production efficiency of this method is significantly higher than that of other hydrogen production methods coupled with PWR. The hydrogen production efficiency of this method can be higher than 42%, which is more than 50% higher than that before optimization.
The comparison of the hydrogen production efficiency of this method with that of other renewable energy is analyzed. The result shows that the hydrogen production efficiency of this method is more than 20% higher than that of renewable energy.
This work could improve the efficiency and energy utilization of PWR to hydrogen production in China, which could promote nuclear energy to help decarbonized and mitigate climate warming.
Research on the Applicability of the IWF Model to Coupled Heat Transfer Between Intra- and Inter-Assembly in LFR
Authors: Wu Di , Gui Minyang , Cheng Jie , Wang Jianjun 
Published: November 2024

Abstract
As one of the Gen-IV advanced nuclear reactors, the lead-based fast reactor (LFR) has high power density, strong inherent safety, and excellent maneuverability, which makes it easy to realize modularization and miniaturization and can be used for conventional power production and energy supply needs of remote areas. The premise of the engineering application of LFR is to ensure the safety and reliability of reactor operation, the research of thermal-hydraulic and safety characteristics is quite important for the design and optimization of LFR.
In this study, a numerical analysis method was used to analyze the coupled heat transfer characteristics between intra- and inter-assembly in LFR. Two different models, the multi-layer wall model and the IWF sub-channel model, were studied to characterize the effects of the inter-wrapper flow (IWF). These two models were validated and compared based on the IWF experiment carried out by German KIT with the THESYS experimental platform. The results show that the accuracy of the IWF sub-channel model is higher under different working conditions, and the deviation between the fluid temperature in the inter-assembly and the experimental value is less than 2%. The calculation accuracy of the multi-layer wall model is acceptable for the low IWF flow rate, and the deviation increases gradually with the IWF flow rate. Based on the IWF sub-channel model, the axial fluid temperature distribution in IWF was further analyzed under different non-uniform boundary conditions.
JENDL-5 Benchmarking for Advanced Test Reactor for Preparing Burnup Analysis Using Isotopic Data From HTGR Type Fuel Irradiation Tests
Authors: Okita Shoichiro , Aoki Takeshi , Fukaya Yuji , Tachibana Yukio 
Published: November 2024

Abstract
Japan Atomic Energy Agency (JAEA) has been developing a methodology for nuclide generation-depletion and decay heat evaluations for High Temperature Gas-cooled Reactors (HTGRs). We are planning to perform validation of the evaluation method with isotopic composition data obtained from HTGR type fuel irradiation tests (AGR tests) performed at the Idaho National Laboratory. As a first step of this plan, preliminary validation of a core calculation model of the Advanced Test Reactor (ATR) is performed with a continuous-energy Monte Carlo code MVP-3. We made the calculation model with the latest nuclear data library in Japan called by JENDL-5. We also calculated effective multiplication factors and relative power densities for the ATR calculation model. As a result of comparison with measured values reported in the IRPhE handbook, the JENDL-5 and the calculation model built with MVP-3 showed a good calculation accuracy within experimental error range in the effective multiplication factor and ±8.8% in the relative power density distribution. The results will help us to perform our planned validation of our nuclide generation-depletion and decay heat evaluation methodology with the AGR test data.
Multiple Knowledge Exploration Functions for Advanced Reactor Design and Safety in Knowledge Management System Implemented in the ARKADIA
Authors: Seki Akiyuki , Kondo Yuki , Hashidate Ryuta , Yoshikawa Masanori , Yokoyama Kenji , Takaya Shigeru , Enuma Yasuhiro , Hazama Taira , Wakai Takashi , Asayama Tai 
Published: November 2024

Abstract
Japan Atomic Energy Agency (JAEA) is developing ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) to meet the expectations for safe, economical, and sustainable carbon-free energy. The Knowledge Management System (KMS) included in ARKADIA consolidates technical information on Sodium-cooled Fast Reactor (SFR) development obtained from “Monju” and other projects. Currently, several search functions are being developed in this KMS according to the characteristics of the information requested by users. First, indexes are added to the documents to quickly find specific data in KMS through the portal site. In addition, a search function has been enhanced to enable cross-search of all information registered in KMS. For information regarding facilities, an enhancement of the search function for Computer Aided Design (CAD) drawings and point cloud data has been introduced. With this function, confirmations that used to require on-site visits can now be handled remotely. Furthermore, the possibility of information retrieval using knowledge graphs is being developed to enhance efficiency of information retrieval, along with the evaluation flows in ARKADIA. This function can integrate rules, standards, interpretations, guidelines, design information, and related feasibility check information into a knowledge graph. Finally, milestones in KMS development, integration with the other two systems, and further enhancements are presented.
Development of an Integrated Point Kernel Shielding Calculation Code for Fast Three-Dimensional Radiation Field Characterization
Authors: Chen Junyi , Li Ruihan , Liang Jingang 
Published: November 2024

Abstract
Accurate and efficient shielding calculation code is crucial for radiation protection and safety of nuclear facility. Poker, a new integrated point kernel shielding code, has been developed for fast three-dimensional radiation field characterization. Point kernel method, Construct Solid Geometry (CSG) representation method, and various appropriate multi-layer buildup factors (BUFs) formulas are integrated into the code. Verification and validation of the code have been carried out by benchmark problems available in the literatures. The code is validated by estimating BUFs, MLBUFs, and gamma dose rates at the exit of the shield with standard database and experiments. Further investigation of the code’s extensivity in handling deep shielding geometries arrangements of up to 100 mean free paths (MFPs) is also verified by comparing it with the results of previous work. The results demonstrate the accuracy of the code in deep penetration scenarios. Moreover, analysis is carried out to identify characteristics of the radiation dose rate in the typical simple volumetric source term scenarios, and comparison of this code with existing point kernel code is carried out. This code is written in C++ language and is compatible with MacOS, Windows, and Linux operating systems.
Thermal Hydraulics Simulation of a Typical Pressurized Water Reactor Coolant System Using CFD Method
Authors: Zhang Mingqian , Lin Run 
Published: November 2024

Abstract
The thermal hydraulics simulation of the reactor coolant system for a typical three-loop pressurized water reactor was conducted based on numerical solution of Reynolds-averaged Navier-Stokes equations by commercial CFD software. This study aims to obtain the three-dimensional, global and localized flow features of the reactor coolant system. The completed model of the reactor coolant system is built including reactor pressure vessel and internals, core, steam generator, primary pump and linking pipe. The flow and the temperature have been investigated under normal steady operating condition with the full core thermal power and unbalanced operating condition with the failure of any primary pump during cold shutdown state. The local thermal hydraulic features of the reactor pressure vessel head dome, the thermal stratification in the reactor pressure vessel upper plenum and the hot legs, and the swirling flow of the primary pump are characterized to give reference to the reactor safety operation. This analysis practice provides an effective evaluation for the three-dimensional thermal hydraulics phenomena, and these encouraging results allow performing the comprehensive analysis for the reactor coolant system.
Application of Modal Decomposition Techniques for Reduced-Order Modelling of Once-Through Steam Generator
Authors: Xu Yifan , Peng Minjun , Xia Genglei 
Published: November 2024

Abstract
A fast simulation model that provides accurate estimation for thermal hydraulic parameters of the Once-through Steam Generator (OTSG) is of great significance for nuclear safety and economics, and can greatly reduce the computational cost of high-fidelity models for performing multi-query tasks. The Dynamic Mode Decomposition (DMD) and Proper Orthogonal Decomposition (POD) are commonly used non-invasive model reduction techniques and are data-driven which means the methods generate low-dimensional models as approximate representations of Full Order Model (FOM) rely only on data, thus have more advantages in reduced-order modeling. While the applications of these methods are rarely involving the OTSG. This paper aims to analyze the applicability of DMD and POD for OTSG reduced order modelling. The simulation data of main parameters calculated via RELAP5 are collected as snapshots. The influence of snapshots and modes truncation on the estimation performance of ROMs is analyzed. The results show that the POD model outperforms the DMD model when the same number of modes are used to build the ROM, although the energy ranking method is introduced to extract the main modes to improve the accuracy of the DMD model.
Data Augmentation for Discharging Time Prediction of Particle Flow: A Deep Learning and Style Transfer Approach
Authors: Wu Mengqi , Liu Yang , Li Bin , Zhang Zhen , Gui Nan , Tu Jiyuan 
Published: November 2024

Abstract
Recent years have witnessed an increasing interest in utilizing deep learning models trained on large datasets for particle flow research. The discharging time of particle flow characteristics can be well predicted by the deep learning methods. However, the collection of a substantial amount of experimental data (real data) can be burdensome due to the significant manpower and equipment expenses involved. To address the challenge of small and insufficient samples, we have developed a data augmentation strategy with transfer learning that effectively combines real and simulated data, thereby synthesizing data to expand the dataset at a minimal cost. In this specific study focusing on discharging time prediction, we generate synthetic data that combines the flow characteristics from simulated data with the image style from experimental images since the time required for simulation and synthetic data generation is negligible compared to the effort involved in obtaining real data. The results indicate that the style transfer model can successfully generate synthetic data with realistic styles, significantly reducing data collection burden. Besides, models trained with synthetic data have the ability to predict discharging time for real scenes images of particle flow. Our proposed strategy provides a feasible data augmentation approach for particle flow research, and further efforts are needed on data quality and network capabilities in the future.
Pre-Test Analysis of a PLOFA Scenario for CIRCE-THETIS Facility by Means of Coupled STH/CFD Calculations
Authors: Stefanini Pietro , Pucciarelli Andrea , Forgione Nicola , Di Piazza Ivan 
Published: November 2024

Abstract
In the context of GEN IV fission power reactors, the Liquid Metal Fast Reactors (LMFRs) stand out as a promising solution, introducing a better radioactive waste management and new safety features with respect to previous NPPs generations. Numerous EU projects have been started to support the development of such kind of reactors. This work is in support of the CIRCE-THETIS experimental campaign foreseen at the ENEA Brasimone Research Centre (RC), in the framework of the EU project PATRICIA.
This paper focuses on the pre-test analyses of one of the postulated Protected Loss Of Flow Accident (PLOFA), foreseen during the experimental campaign, investigating the safe cooling strategies for such kind of scenarios: in the considered case, the SG was used as decay heat removal system. The facility was represented using a coupled CFD/STH approach with an explicit time advancing scheme. The model was deeply examined to ensure the best representation of reality, aligning with the preliminary evaluation from the experimentalists. Particular attention was paid to the SG heat transfer prediction due to the new helical geometry that provides heat transfer enhancement and to the thermal-hydraulics phenomena occurring inside the pool during the postulated transients.
Development and Initial Analysis of a Neutron Diffusion Model for TEPLATOR Using COMSOL Multiphysics Software
Authors: Ray Dipanjan , Lovecký Martin , Závorka Jiří , Škoda Radek 
Published: November 2024

Abstract
The growing demand for sustainable energy solutions has stimulated the advancement of novel reactor technologies like TEPLATOR, a small modular reactor conceptualized by a collaborative team of researchers from the University of West Bohemia (UWB) and Czech Technical University (CTU). TEPLATOR utilizes a heavy water-cooled and moderated pressure channel-type system, providing a viable alternative to fossil fuel-based plants for industrial and district heating requirements. This assists in minimizing pollution and mitigating environmental consequences. This article covers the initial stages of developing a neutron diffusion model for the TEPLATOR core using COMSOL, a finite element-based multi-physics software package. The study focuses on creating and analyzing two- and three-dimensional diffusion models to understand spatial neutronic behavior and determine integral reactivity worth under various control rod positions. Monte Carlo-based neutron transport code Serpent is employed to compute neutronic parameters, incorporating the integration of multi-group constants into the diffusion model. Steady-state calculations for the Beginning-of-Cycle condition (BOC) are emphasized, with a comparison between computed values from the COMSOL diffusion model and Serpent simulations, along with the presentation of neutron flux distribution profiles.
Dynamic Refinement Evaluation and Criteria Optimization for Flow Features in Preparation for LES of Turbulent Hydrogen-Air Deflagration Experiment
Authors: Venckus Julius , Povilaitis Mantas 
Published: November 2024

Abstract
One of the worst case scenarios in a nuclear power plant is a hydrogen explosion. The explosion strength depends on turbulence, as it may stretch and accelerate the flame, therefore enhanced understanding of flame-turbulence interaction is crucial for accurate explosion risk assessment. LES is required to numerically study flame-turbulence interaction in a practically relevant scales. Unfortunately, larger scale LES even with dynamic mesh refinement could require substantial computational resources. To determine optimal sufficient mesh and dynamic refinement criteria to resolve both flow structures and flame without performing full-scale calculations, it is desirable to first carry out small-scale simulations. In this study mesh requirements sufficient to resolve flow features corresponding to the conditions more extreme than in a reference large-scale experiment were investigated. Simulations were carried out using open-source CFD toolkit OpenFOAM-11. Several static and dynamic meshes were evaluated by comparing both fraction of resolved kinetic energy and maximum mean downward velocity, as coarse meshes failed to correctly resolve transitions to supersonic flows. For refinement criterion, velocity gradient-dependent multiplier from WALE (Wall-Adapting Local Eddy-viscosity) SGS (Subgrid-scale) model was chosen. It was determined that 8 mm base mesh with 2 levels of dynamic refinement with criterion value 2500 s−0.5 resembles static 2 mm mesh results well and is enough to resolve dominant flow features in the domain. The study thus serves as a preparation for the full-scale LES simulations.