bluuebunny/crossref_metadata_2025 huggingface dataset:
Multiphysics Coupling Analysis of an FCM-Fueled Gas-Cooled Microreactor
Authors: Hu Jipu , Shen Yuyang , Wang Ruixiang , Feng Kuaiyuan , Lou Lei , Guo Hui
Published: November 2024

The micro high-temperature gas-cooled reactor (mHTGR) is a compact, efficient, and safe nuclear reactor using fully ceramic microencapsulated (FCM) fuel. This study focuses on the core design and multi-physics analysis of a 15MWt mHTGR, which uses helium as the coolant, graphite as the moderator and reflector, and control rods for reactivity control. The reactor is designed to operate continuously for 20 years without refueling. A model coupling multiple physics of a typical fuel assembly was developed using OpenMC and ABAQUS software. The model is simplified by dividing it into 3D solid heat transfer with 2D fluid-solid interface conduction and 1D fluid heat transport. Coupled multi-physics simulations are then performed under steady-state and two typical accident conditions. During steady-state operation, the core coolant flow rate is 9.0 kg/s, and the outlet temperature is 904.6 K. The maximum temperature of 1149 K occurs at a distance of 100 cm from the outlet. In transient simulations, the peak fuel temperature reaches 1204 K during a 200 pcm reactivity insertion and remains almost unchanged during a loss of flow accident. The results show that the core exhibited good transient response characteristics and inherent safety.
The Investigation of Heat Transfer Deterioration in Supercritical Water Heating Systems
Authors: Chen Yu-Sen , Lee Jin-Der , Chen Shao-Wen
Published: November 2024

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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.
The Steady States and DLOFC Accident Analysis of HTR-PM in the OTTO Operation Scheme
Authors: Liu Songyang , Li Xuelin , Luo Yong , Liu Wei , Zhou Qin , Gu Cheng , Zhang Tian
Published: November 2024

High Temperature gas-cooled Reactor – pebble bed module (HTR-PM) located in Shidao Bey, Shangdong province, China, which is the first commercial HTR in the world, adopts spherical fuel element with TRISO particle and graphite material as moderator. As the national demonstration project of China, the outlet coolant temperature of HTR-PM can reach 750 degrees and more than 550 degrees of steam, this reactor can be used not only in electrical power generation, but also in industrial heat source supply, hydrogen production and petrochemical industries. Besides, the module design can decrease the financial cost during the construction period, and shorten the time in manufacturing.

For the online reloading operation, the MEDUL scheme was adopted in HTR-PM operation with 15 fuel batches. To shorten the running-in phase and simplify the reloading system, the OTTO scheme can be adopted to replace MEDUL in HTR-PM operation, in this paper, the steady states calculation based on VSOP-THERMIX has demonstrated the neutronic and thermal hydraulics parameters of HTR-PM in the OTTO scheme from initial critical situation to equilibrium states.

With different assumed OTTO operation schemes, the spatial displacement of max power density and max fuel temperature in the core is shown in every loading step. Meanwhile, due to the moderation ability of graphite, the isotope content of uranium and plutonium compared with different loading methods.

Last, based on the coupled calculation of equilibrium states, the maximum fuel temperature analysis of DLOFC in every loading stage by THERMIX demonstrates HTR-PM with OTTO scheme still keeping inherent safety features. The results show the fraction of fuel elements in different loading stages has an effective influence on maximum fuel temperature.
Optimization of Inventory Control Strategy for Part Load Operation in a sCO2 Recuperated Brayton Cycle for a 5 MWth Heat Pipe Micro Modular Reactor
Authors: Peiretti Matthias , Hofer Markus , Buck Michael , Meucci Ruggero , Starflinger Jörg
Published: November 2024

This study focuses on the implementation of an inventory control strategy for part-load operation in a 5 MWth Recuperated supercritical carbon dioxide (sCO2) Brayton Cycle, integrated with a Heat Pipe-Cooled Micro Modular Reactor (MMR). A preliminary investigation emphasises the crucial role of the inventory tank design for efficient control.

However, when relying solely on the inventory tank for control, matching lower power demands is challenging due to the low pressure at the compressor inlet. Without countermeasures, subcritical conditions are reached when the power decreases below 43% of the nominal capacity of the studied system. To overcome this limitation, a combined control strategy which involves increasing the inlet temperature of the compressor is investigated. This allows to reach 8% of the nominal power.

The study further investigates the performance of the system under load-following conditions. An aggressive power rate change of 7.5%/min of the full load until 30% of the nominal power is simulated, with the resulting power output closely tracking the demand.
Optimal Design of Supercritical CO2 Power Cycle for High Temperature Gas-Cooled Reactor
Authors: Zhou Yujia , Zhang Yifan , Li Hongzhi , Yao Mingyu
Published: November 2024

The supercritical carbon dioxide (sCO2) Brayton cycle has been regarded as a promising technique for future energy conversion systems. sCO2 recompression cycle configuration has been widely studied for nuclear reactors with the temperature difference between the reactor inlet and outlet restricted to 200°C. However, this cycle configuration may be not suitable for high temperature gas-cooled reactors (HTGR) with a temperature difference up to 500°C. Herein, we aim to develop an optimal sCO2 cycle configuration for HTGR with a broad temperature span, which has both high cycle efficiency and sufficient mechanical security guaranteed by energy cascade utilization. We propose a recompression cycle combined with reheat and expansion technologies, and perform a parameter optimization in Fortran platform. First, the thermodynamic models of the system are established. Then, the parametric analysis is conducted about the effects of split ratio, temperature difference of the heat exchangers, inlet parameters of turbines and compressors, and isentropic efficiencies on the cycle performance. The sCO2 cycle tailored for HTGR can achieve a cycle efficiency of 44.98%. Our work not only highlights the design criteria of temperature-match between the HTGR and sCO2 cycle but also guides the numerical assessment of the steady-state performance.
Construction and Exploration of a Diagnostic Platform for High-Temperature Gas-Cooled Reactor Units
Authors: Mao Cui , Liu Haisheng , Geng Di , Liu Yuhua , Yin Zaizhe
Published: November 2024

The high-temperature gas-cooled reactor (HTGR) is an advanced fourth-generation nuclear reactor technology known for its advantages in safety, efficiency, economic viability and versatile applications. The secure and stable operation of HTGR is crucial for ensuring sustainable energy supply. The HTGR unit diagnostic platform integrates functions such as data collection, fault recognition and predictive analysis to provide comprehensive support for the operation and maintenance of HTGR. This paper begins by reviewing the background and development of HTGR technology and subsequently discusses the architectural design, key technologies and practical application effect of the diagnostic platform. Through the development and utilization of this platform, operators can monitor the HTGR status in real time, accurately predict potential faults, and thereby ensure its safe and efficient operation.
Validation of Doppler Reactivity Feedback in SPERT-III E-Core With the Best-Estimate Transient Code TRAC Toshiba Version
Authors: Egawa Tohru , Tokashiki Mikio , Miyaji Takamasa , Fukunaga Takanori
Published: November 2024

We have developed a best-estimate transient code TRAC Toshiba version, which is Toshiba’s advanced version of the TRAC code for BWR. This code is expected to simulate thermal hydraulic and neutron kinetics phenomena three-dimensionally with high accuracy for BWR transient events. For the validation of the code, reactivity accident tests in cold-startup of The Special Power Excursion Reactor Test III (SPERT-III) E-core were calculated and compared with experiment results. Two lattice code “NEUPHYS” and “TGBLA” were used for the calculation. “NEUPHYS” developed by Nuclear Fuel Industries, Ltd. is a neutron transport calculation code with Method of Characteristics. “TGBLA” is a current licensing code of Toshiba. For the confirmation of properly simulation of initial core condition of the tests, static nuclear characteristics of the core were calculated. keff, spatial power peaking, ∧/βeff, worths of control rod and transient rod were compared with experiment results, and good agreements were confirmed. In the transient tests, estimated C/E-1 of peak power without simulation error in transient rod worth were −2%±4% for “NEUPHYS”/TRAC and 1%±4% for “TGBLA”/TRAC in high reactivity insertion tests more than 1$ excluding test 20 which is considered outlier case (±4% were evaluated from uncertainty of experiment). In high reactivity insertion transient tests, temperature of UO2 pellet behaves adiabatically by the time of power peak. Therefore, temperatures of UO2 pellet at the power peak are expected to contain small amount simulation error, it is expected that doppler reactivity simulation error are dominant in aforementioned simulation errors.
Study on Fluid-Elastic Instability of Tube Bundles in Cross Flow Based on Spatiotemporal Coherence
Authors: Cheng Yuxuan , Bu Shanshan , Kai Guo , Chen Deqi
Published: November 2024

The flow-induced vibration (FIV) of tube bundles of steam generator is a complex fluid-structure interaction with coupling of multibody. With the impact of the surrounding tubes, the vibration of the tube is excited by forces with different frequencies, and the vibration of tube bundles turns to be nonlinear. At the same time, due to the existence of asymmetry in the support of the tube bundle, resulting in different intrinsic frequencies in the streamwise and transverse directions of the tube bundle, which can result in the difficulty in perdition of fluid elastic instability (FEI). In this work, the fluid elastic instability of tube bundle with different frequency ratios Rf is investigated through water tunnel experiment and numerical simulation. The target tube arrays are rotated triangular arrays with pitch ratio of 1.33 and 1.41. The effects of frequency ratios on the amplitude and frequency response of the tube bundle are analyzed, while the coherence is induced into the analysis of the route to instability. The results show that the critical velocity can be affected by frequency ratios Rf, but the effect of Rf differs in tube arrays with different pitch ratios. The evolution of Lyapunov exponents shows the nonlinearity of the vibration can be influenced by Rf. The method and results of this work can be helpful in the prediction critical velocity of FEI in tube bundles in cross flow.
Dynamic Modeling and Characteristic Analysis of sCO2 Brayton Cycle System
Authors: Yun Shichang , Zhang Dalin , Li Xinyu , Lv Xindi
Published: November 2024

The supercritical carbon dioxide Brayton cycle has advantages such as high thermal efficiency, inherent safety, simple structure, and compact space at medium to high heat source temperatures. Being able to cooperate with nuclear energy, solar energy, fossil energy or other heat sources for power generation has enormous development prospects. The FUSTAR system proposed by Xi'an Jiaotong University, utilizing the high core outlet temperature of the fluoride salt reactor and the supercritical carbon dioxide Brayton cycle system, has the characteristics of high efficiency, high safety, compact structure, and anhydrous cooling. It proposes a new solution for power supply in remote areas. However, the SCO2 Brayton cycle system is a highly integrated system, and its dynamic characteristics are very complex. There is an urgent need to establish a dynamic model of the SCO2 Brayton cycle system to lay the foundation for subsequent control system design. Based on the conservation equations of mass, momentum, and energy and the thermal hydraulic model, heat exchangers, regenerators, and coolers were established. A lumped parameter calculation model for turbine compressors has been established. A SCO2 Brayton cycle system model was built based on the MATLAB/Simulink platform, and the dynamic response of each device under disturbance was calculated. The results indicate that there are significant challenges in the stable operation of the SCO2 Brayton cycle under the most efficient operating strategy. The inlet parameters of compressors and turbines are highly sensitive to the outlet fluid pressure and temperature, and other auxiliary equipment is needed to ensure relatively stable inlet conditions for turbomachinery. The thermal inertia of the heat exchanger is relatively large, and the control response of the SCO2 Brayton cycle system requires a long time. The system responds slowly to changes in turbine and compressor speed and cannot be quickly adjusted. Rapid adjustment devices such as turbine throttle valves are required to respond quickly to changes in system load. The modeling and analysis method of SCO2 system can also be used for other configurations.
An Experimental Study on the Scrubbing Characteristics of Graphite Dust Aerosol in HTGRs
Authors: Wang Yating , Zhang Yiyang , You Zhikai , Zhao Shumiao , Fang Zhu , Sun Libin
Published: November 2024

The potential release of graphite dust of HTGR contributes the most to the source term and has a great impact on the safety of reactors. In water ingress accidents, the graphite dust may be released to the containment during the overpressure discharge, which not only contributes to the source term but also makes trouble for post-accident measures. An important improvement for reducing the radioactive material emissions in the light water reactor (LWR) is the aerosol pool scrubbing. The effectiveness of pool scrubbing with sphere particles has been extensively investigated. whereas, there is little research on the passive removal of dust and the source term for HTGR. This paper explores the possibility of pool scrubbing retaining the graphite particle and fission products in the overpressure discharge. The scrubber system is established and the pool scrubbing experiments with flake graphite particles and spherical silica particles are conducted. The effects of various parameters on graphite scrubbing efficiency are investigated, including water submergence depth, gas flow rate, particle size, and particle mass concentration. The decontamination factor (DF) is a key parameter to describe the scrubbing efficiency. It is shown that the DFs increase significantly with the increase of submergence depth and gas flow rate both for two kinds of particles. Dominated by the inertial deposition mechanism, the decontamination efficiency of graphite is significantly lower than silica particles for particles larger than 1 μm. As for the particles smaller than 1 μm, the difference in decontamination efficiency between graphite and silica is small attributed to the Brownian diffusion mechanism. In general, the total DF of silica is significantly higher than that of graphite particles. To improve the retention efficiency of particles, a metal packing breaking element is placed in the vessel. It benefits the breakup of large globules and the contacting of gas/liquid. In case, the total DF of graphite aerosol increased by an average factor of 2.8 times, meanwhile the pressure drop increased only by 12.4%. It is revealed that almost all particles achieve a DF of 20, corresponding to a retention efficiency exceeding 95%, at the water depth of 110cm. These results are beneficial for the design of scrubber structures that can accommodate low-pressure discharge airflow, which can achieve a high radioactive retention efficiency and maintain a relatively low pressure drop.
Simulation Study of Turbine Trips Without Scram for CPR1000 Nuclear Power Plant Employing MSHIM Control
Authors: Wu Shifa , Chai Yunzhi , Wan Jiashuang
Published: November 2024

In nuclear power plant operations, several signals may trigger the closure of turbine valve, known as turbine trips. It is desirable that these turbine trips would not cause reactor scram to avoid iodine pit and thus save time and work to restart the reactor. This study investigated the turbine trip without scram transients of CPR1000 Nuclear Power Plant (NPP) implementing Mechanical Shim control system. The Nuclear Steam Supply System (NSSS) consisting of the one-dimensional reactor core model, the nonequilibrium three-region pressurizer model, the nonlinear dynamic model of U-tube steam generator was first modeled. The Balance of Plant (BOP) model including the turbine model, the condenser model, the reheater model and the deaerator model was then established. Based on that, the full-scope simulation platform of CPR1000 NPP was constructed with relevant control systems integrated on MATLAB/Simulink. Meanwhile, the protection system of CPR1000 was also considered in which different real-time protection signals will be collected and processed, and once there is one or more signals were triggered, the protection system will be triggered and shut down the reactor for safety reasons. Finally, the most serious turbine trip transient from 100% Full Power (FP) were initiated and simulated. Results showed that this CPR1000 NPP configured MSHIM control can withstand the above transient as well as provide enough safety margins without leading to a reactor scram.
Assessment of Emergency Planning Zone for Small Modular Pressurized Water Reactors Based on the Risk-Informed Methodology
Authors: Wang Mengxi , Wu Nan , Cao Qun , Yan Jin , Gao Jiaxuan , Zhang Jiemin , Xue Na
Published: November 2024

More than 80 Small Modular Reactors (SMRs) designs are under development worldwide. SMRs usually have the characteristics of integrated design, modular manufacturing and installation, high safety, and proximity to users, and have broad prospects for development and application. Due to the lower power levels and passive safety features of SMRs, the radioactive releases, both in magnitude and frequency, would be significantly lower than that of large reactors. Too large emergency planning zone (EPZ) size may limit the flexible deployment of SMRs. How to reasonably determine the EPZ size of small reactors has become a focus in the nuclear industry. Based on the risk-informed methodology recommended by IAEA-TECDOC-1652, considering both the frequency of the entire spectrum of accident scenarios and the dose consequence for each scenario, this paper analyzes and calculates the EPZ size of a certain small modular pressurized water reactor in China using the Level-2 PSA (Probabilistic Safety Analysis) source terms and the site actual meteorological data. Based on the realistic models, the preliminary recommended EPZ size is about 700m. Even in the conservative meteorological conditions, the EPZ size is about 2.7km, which shows a significant reduction than that of large reactors. The risk-informed methodology is an application of Level-3 PSA supporting in the EPZ sizing. The results could further improve the public’s acceptance and recognition of SMRs and promote their safe and reliable deployment worldwide.
Development of Two-Dimensional Heat Conduction Model in Subchannel Code Based on OpenFOAM
Authors: Dong Zhengyang , Liu Kai , Wang Mingjun , Wu Junmei , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: November 2024

Subchannel codes play an important role in the thermal-hydraulic analysis of reactor cores. Based on the open-source CFD software OpenFOAM, the research team developed CorTAF, a finite volume method-based thermal-hydraulic analysis code for the core subchannel level. All the CorTAF codes are done on OpenFOAM, which has the function of calculating different phenomena in different reactor types. Compared with the traditional subchannel codes, CorTAF utilizes the advanced discrete algorithm of OpenFOAM to distinguish different plane directions on the flow, and implants more advanced models to calculate the phenomena in the reactor core, which has many advantages such as higher computational accuracy, faster computational efficiency, and better developability. Most of the subchannel codes ignore the axial heat transfer when calculating the temperature of the fuel rod, and the corresponding error is introduced here. In this paper, the axial heat conduction of fuel rods is further studied on the basis of CorTAF thermal-hydraulic model. On the basis of the previous model which only considers the heat conduction of radial nodes, the axial heat conduction is considered, the construction and solution of two-dimensional heat conduction matrix is realized, and the two-dimensional unsteady heat conduction model of fuel rod is developed.

The developed two-dimensional unsteady heat conduction model is used to calculate the normal and blocking conditions respectively, and compared with the previous node heat conduction model, it is found that: Under normal operating conditions, there is little difference between them, and the maximum error is less than 0.5K. Under the condition of blocking accident, the overall temperature distribution trends of the two models are basically the same, but the temperature distribution of the two-dimensional heat conduction model is more uniform, and the maximum temperature of the fuel pellet is smaller. In a word, the upgrade of CorTAF fuel rod heat conduction model further improves the accuracy of core thermal-hydraulic calculation, and provides support for its calculation of special working conditions such as blockage accident, which can be used for reference in similar program development.
Development of DLOFC Accident Analysis Function of DAYU
Authors: Dou Haoming , Lang Minggang , She Ding , Zhang Han
Published: November 2024

As a typical design basis accident (DBA) of High Temperature gas-cooled reactor (HTR), the Depressurized Loss Of Forced Cooling (DLOFC) accidents which will result in higher peak temperature receive more and more attention. The dynamics code for accidental system utility (DAYU) for HTR is developed by the Institute of Nuclear and New Energy Technology (INET) to conduct reactor accident analyses. DAYU is designed for thermal design and steady-state performance study, so the change of system pressure caused by the reduction of coolant inventory in the primary circuit can’t be characterized. In this work, the loss mass of coolant is introduced into the system pressure equation to characterize the pressure change process. In effect, the mass loss rate caused by the break will not only lead to the decrease of system pressure but also affect the flow direction of coolant in the primary circuit due to the different locations of the break, which will result in different temperature of the coolant that leaks out. Taking this phenomenon into consideration, the break mass flow rate is covered in the core flow field calculation further, which makes the simulation of the change of flow field in the reactor caused by the break mass flow rate become a reality. With these improvements, the DLOFC accidents analysis function of DAYU has been accomplished.
Time Series Analysis With Combined Learning Approach for Anomaly Detection in Nuclear Power Plants
Authors: Dong Feiyan , Chen Shi , Demachi Kazuyuki , Yoshikawa Masanori , Seki Akiyuki , Takaya Shigeru
Published: November 2024

Timely and accurate anomaly detection is of utmost importance in ensuring nuclear safety to guarantee the normal operation of nuclear facilities and to prevent severe accidents. Due to the complexity of the dataset from the condition-based monitoring, advanced data-driven deep learning algorithms with their state-of-the art performance, are deployed for automatic anomaly detection in nuclear power plants (NPPs). Nevertheless, in the collected dataset, the knowledge about whether some certain periods are in an abnormal or normal state is available, while for a greater portion of the time, only data in unknown operating states can be acquired, rendering the general supervised learning (SL) and unsupervised learning (UL) suboptimal learning methods. Therefore, a combined learning approach is deployed in the proposed time series analysis framework, allowing the utilization of additional domain knowledge within the dataset to train the framework from SL to UL seamlessly. Additionally, due to the intricate inherent relationships among different instruments and the low frequency of anomaly occurrences, the implicit influential weights of each dimension vary when identifying different kind of anomalies. To tackle this challenge, an attention mechanism is integrated into the anomaly detection model for highlighting the prominent abnormal features and hence enable it to address the subtle impact of each dimension to enhance the ability of anomaly detection task. Experimental evaluations on the real-world multivariate benchmark dataset Security Water Treatment (SWaT) and on the high temperature gas-cooled reactor (HTGR) anomaly cases dataset based on the analytical code ‘ACCORD’, demonstrate the feasibility and effectiveness of the proposed approach.