bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of the AI-Assisted Thermal Hydraulic Analysis Method for Condensing Bubbles in Vertical Subcooled Flow Boiling
Authors: Zhou Wen , Miwa Shuichiro , Okamoto Koji , Okawa Tomio , Tsujimura Ryoma , Nguyen Thanh-Binh
Published: November 2024

Subcooled flow boiling plays a pivotal role in various industrial applications, including nuclear reactors and thermal management systems. However, the rapid and complex changes of condensing bubbles experience from their inception to collapse in subcooled flow boiling present substantial obstacles for conventional bubble detection methods, especially when it comes to condensation bubbles. In light of this, a state-of-the-art AI method is developed and validated for the efficient detection and tracking of condensation bubbles in subcooled flow boiling, thereby enabling the effective execution of thermal hydraulic analyses. This study initially employs computer vision technology to efficiently construct a bubble dataset. Subsequently, a bubble detection model is trained based on this dataset and the YOLOv8 with attention mechanism. Utilizing the results of bubble detection, a multi-object tracking algorithm is then applied to track all bubbles in each frame. The developed AI-based method exhibits a strong ability to detect 95% of condensation bubbles in subcooled flow boiling, additionally, it streamlines the extraction of critical thermal hydraulic parameters, including void fraction, bubble lifetime, and nucleation site density. The model’s accuracy and consistency is demonstrated compared to empirical correlations, affirming its reliability in analyzing the intricate dynamics of subcooled flow boiling.
A Novel Optical Chamber Design and Shielding Optimization for Application in Liquid Scintillation Counting
Authors: Wang Husheng , Zuo Liangzhou , Yan Yangyang , Jiang Hao , Wang Jie
Published: November 2024

Liquid Scintillation Counting (LSC) is a technique widely employed for measuring the activity of radioactive isotopes, finding applications in fields such as radioactive isotope measurements, nuclear material analysis, environmental monitoring, and nuclear medicine. This study is based on the Triple-to-Double Coincidence Ratio (TDCR) method within liquid scintillation counting, proposing a novel TDCR-LSC detector structure with a TriL-shaped optical chamber. Compared to traditional TDCR structures, the TriL-shaped optical chamber integrates the U-shaped design from the Pacific Northwest National Laboratory (PNNL) and the conventional TDCR structure. This is achieved by cutting three L-shaped hollow light guide tubes in the shielding, directing the light from the liquid scintillation vial at a 90° angle into three vertically positioned photomultiplier tubes (PMT). The design avoids the need for heavy external shielding around the PMT, and the L-shaped light guide prevents external radiation from directly entering the sample, thereby enhancing the shielding performance. The TriL-shaped design exhibits significantly improved shielding efficiency compared to the traditional TDCR structure. Validation through Geant4 simulations demonstrates that the TriL-shaped design outperforms the traditional TDCR structure in terms of shielding efficiency while maintaining efficient light collection. This research introduces a novel design approach to enhance the performance of liquid scintillation counting detectors.
Modeling Co-Current and Counter-Current Flow: A Performance Evaluation of the TRACE Condensation Model with Non-Condensable and Light Gases
Authors: Albdour Samah A. , Addad Yacine , Afgan Imran
Published: November 2024

Efficient steam condensation is crucial for ensuring the secure functioning of nuclear power plants (NPPs) by mitigating the potential dangers associated with excessive pressure and overheating. Nonetheless, the existence of non-condensable gases (NCGs) can obstruct this essential process, forming thermal resistance layers, impeding steam diffusion, and hindering condensation on the system’s surfaces. Consequently, the objective of this study is to enhance our comprehension of steam condensation by assessing its effectiveness in the presence of non-condensable gases. The current work utilizes a condensation setup in KAIST’s passive containment cooling system (PCCS) facility, chosen for its validation accessibility. Unlike prior validations focused on pure steam and air using RELAP5/MOD3.2, this study employs TRACE validation to explore the effects of non-condensable gases, including air, nitrogen, hydrogen, and helium. The research generates diverse scenarios, correlating them with TRACE parameters to understand the impact of gases. Using the user-friendly Symbolic Nuclear Analysis Package (SNAP) graphical interface, a geometric model is built and verified in the TRACE code, aligning well with experimental data. Subsequently, 2,100 TRACE cases are generated to evaluate the condensation heat transfer coefficient (HTC). Pearson correlation coefficients from the data highlight that mass fraction has the most significant adverse impact on the HTC, followed closely by the gas type based on molecular weight. Given these notable negative effects, the initial focus is on how mass fraction and molecular weight influence the liquid generation rate in co-current flow condensation. Results demonstrate that a higher mass fraction and molecular weight reduce the liquid generation rate compared to pure steam. Examining their impact on drain mass flow rate in steam/NCG mixture variations reveals their role in triggering counter-current flow limitations (CCFL). Findings suggest that reaching the steam/NCG mixture flow rate threshold stops condensation, initiating CCFL. Moreover, increased mass fraction and a shift to gases with higher molecular weights lead to an earlier onset of CCFL.
C-14 in Geological Repositories: Source Terms and Transport Models
Authors: Qu Zhengzhe , Guo Jingni , Zhang Qi , Wang Yu , Wu Mengjie , Xie Feng , Wei Liqiang , Cao Jianzhu
Published: November 2024

Spent fuel management is one of the key issues for the sustainable development of nuclear energy. The deep geological disposal of radioactive waste is a vital aspect of spent fuel management, aiming to isolate radioactive sources from humans and circulating groundwater as much as possible. Buffer materials, as a crucial part of the engineering barrier system, play an important role in preventing the infiltration of groundwater and the outward migration of radioactive nuclides. However, due to a variety of unforeseen circumstances, there remains a risk of leakage in geological repositories, potentially leading to the radioactive sources penetrating the engineering barrier system and causing environmental damage. Carbon-14 (C-14) is one of the significant radioactive elements in China’s geological repositories. In the event of repository damage, C-14 could enter the biosphere and infiltrate the human body through the food chain. It could pose internal radiation hazards to human tissues and organs, which becomes a major source of collective radiation dose for humans.

In recent years, there has been significant progress in the study of buffer materials for high-level radioactive waste (HLW) geological repositories in China, including the development of GMZ01 type bentonite as a buffer material. Extensive research has been conducted on the basic physicochemical properties and nuclide adsorption characteristics of these materials. However, studies specifically related to C-14 in these repositories are less common. The methods for assessing the inventory of C-14 in the repositories need to be refined. Although some computational models for simulating the migration process of C-14 from the repositories to the environment have been proposed, there is a lack of comprehensive summaries. Additionally, the overall research framework for the migration process of C-14 in buffer materials is not clear.

Therefore, the study of the C-14 source term in geological repositories and its migration and diffusion behavior in bentonite, such as GMZ01 type, is a crucial part of the radiation safety assessment in geological repositories in China. This paper focuses on C-14, which is a typical radionuclide in radioactive source terms in geological repositories, investigates the research progress on the C-14 source term in geological repositories both domestically and internationally, organizes the methods for assessing the inventory of C-14 in repositories and compiles computational models for the migration and diffusion behavior of C-14 in key buffer materials. This paper aims to provide theoretical support for the analysis of the C-14 source term in high-level radioactive waste geological repositories and the study of C-14 migration processes in buffer materials, which could provide an important research basis for the radiation safety evaluation of high-level waste geological repositories in China.
Fault Detection and Location Method for Nuclear Power Plant Main Systems Based on Low-Dimensional Manifolds
Authors: Ai Xin , Liu Yongkuo , Shan Longfei , Gao Jiarong
Published: November 2024

The nuclear power plants have characteristics such as variable operating conditions, many large and complex system, high-dimensional fault data, and complex nonlinearity. Classical fault diagnosis methods such as BP neural networks and support vector machines lack excellent feature extraction capabilities and have limited data learning abilities, making them unsuitable for multi-fault diagnosis tasks in complex, nonlinear nuclear power plants with multiple parameters. To address these issues, this study proposes a multi-fault diagnosis model for the main system of nuclear power plants based on low-dimensional manifolds. Following the concept of distributed diagnosis, the study divides the large and complex main system of the nuclear power plant into several subsystems, transforming the complex multi-system diagnosis task into simpler fault detection and localization tasks for each subsystem. Subsequently, within each segmented subsystem, considering the strong nonlinear coupling relationships among fault features and the difficulty in feature extraction, the study employs a stacked autoencoder method to reduce the dimensionality of fault data and uses the random forest method for classification to achieve fault detection and localization in each subsystem. Through simulation data testing, the fault diagnosis method developed in this study proves to be feasible, providing a reference for addressing the large-scale fault diagnosis challenges in complex nuclear power plants.
Progress on the Development of DAYU3D Code for HTR Thermal-Hydraulic Design and System Analysis
Authors: Zhang Haojie , She Ding , Shi Lei
Published: November 2024

System analysis code is an important tool for thermal-hydraulic design and accident analysis of nuclear reactors. At present, the commonly used system analysis codes for pebble-bed high-temperature gas-cooled reactor (HTR), such as THERMIX and TINTE, adopt two-dimensional (2D) models in the reactor core. A new three-dimensional (3D) thermal-hydraulic design and system analysis code named DAYU3D is being developed by the Institute of Nuclear Energy and New Energy Technology (INET) at Tsinghua University. DAYU3D is able to calculate 3D solid temperature, one-dimensional (1D) multi-batch fuel element temperature, 3D gas flow, and 3D gas temperature. Porous media model is employed to perform the 3D temperature and flow field calculations for the entire reactor. In the pebble bed core, DAYU3D employs 1D thermal conduction model to accurately obtain the temperature distribution inside the fuel element. By comparing calculation results with SANA experimental data, the 3D solid temperature module of DAYU3D in steady state is validated. Also, the three-dimensional and multi-batch calculation capabilities of DAYU3D are demonstrated through the calculation of a sample steady state case for pebble-bed reactor. It is expected that DAYU3D will provide an excellent platform for the research of complex 3D problems and facilitate the precise design of HTR.
Experimental Investigation on the Survival of Cooling Tube in Molten Pool
Authors: Pei Jie , Quan Fengyang , Li Wei , Yuan Yidan
Published: November 2024

An experimental facility was established to verify the structural integrity of the cooling tube under an intense thermal shock formed by the molten corium. The vertical cooling tube specimen made by stainless steel has a diameter of 80mm and a thickness of 10mm, and it is positioned at the center of the test section. The bottom of the tube was connected with a water tank, then the water level and temperature in the tube could be adjusted by changing the relative height of the tank, and the top of the tube was opened to discharge the vapor to the air. The C/K-type thermocouples located at the outer-surface of tube were used to record time that melt interacted with tube, also used to estimate the deep of molten pool. In this study, the simulant material ZrO2 was melted by using the cold crucible technique, then the liquid melt about 70kg was discharged into the test section. The experimental results implies that as soon as the melt contacted with the cooling tube, the ZrO2-crust with the low thermal conductivity was formed around the tube, and it will be significantly decreased the temperature on the inner and outer wall of cooling tube. So, the local maximum temperature in outer wall was near the melting point of stainless steel, the structural integrity of cooling tube was still kept.
Analysis of Mobile Equipment Configuration for Severe Accidents in Nuclear Power Plants Based on Multi-Unit PSA
Authors: Liu Dalin , Wang Jiangguo , Yan Zhen , Wang Shuting , Zhao Ling
Published: November 2024

Multi-unit safety analysis is of great significance for the safety improvement of most nuclear power plants. Meanwhile, the configuration of mobile equipment for severe accidents is diverse in different nuclear power plants, a mature evaluation method is urgently needed in the industry. The configuration evaluation method proposed in this paper can effectively solve the pain point of insufficient theoretical analysis basis and the lack of systematic and effective method when configuring mobile equipment for severe accidents in nuclear power plants. It can help plant owners to make configuration plans according to their own risk conditions avoiding over- or under-allocation, and improve overall safety while reducing procurement costs. The method is basically applicable to all operating reactor types and all nuclear power sites, which has good universality. It is also targeted and specialized for the specific analysis of each nuclear power plants. In addition, a modeling methodology is developed in this paper to build a MUPSA model for level-1 internal events based on existing SUPSA models, which is then applied to the analysis of mobile equipment configuration for severe accidents at a pilot multi-unit nuclear power site.
Numerical Simulation of Subchannel Flow Boiling Using Five-Component Wall Boiling Model and iMUSIG Model
Authors: Luo Hanwen , Wang Hongbin , Xiong Jinbiao
Published: November 2024

Boiling heat transfer of subchannel has significant influence on nuclear reactor thermal-hydraulic performance. Employing the Euler-Euler two-fluid approach in computational multi-fluid dynamic (CMFD) simulations, this study conducts a high-fidelity multi-fluid simulation on subchannels. Utilizing the OECD/NRC NUPEC PWR Subchannel and Bundle Tests (PSBT) benchmark, four subchannels are numerically simulated: a typical central subchannel, a central subchannel with a thimble, a side subchannel, and a corner subchannel. The simulation model integrates a five-component wall boiling model and an inhomogeneous multiple size group (iMUSIG) model in OpenFOAM v6, capturing phenomena like bubble sliding, size variability, coalescence, and breakup. Turbulence is modeled using the standard k-ϵ and Sato models, the latter incorporating an additional turbulent viscosity term for bubble-induced turbulence. The study compares cross-sectional averaged void fractions at measurement planes and analyzes key bubble dynamics parameters, including bubble departure and lift-off diameters, wall heat flux partitioning, and bubble size distribution. The results have shown the promising accuracy of the five-component wall boiling model coupled with the iMUSIG model in predicting the void fraction.
Simulation Study of Transient Characteristics of Liquid Lead-Bismuth Natural Circulation Loop
Authors: Sheng Zhenhua , Yang Zhen , Geng Tengjun , Wang Shengfei , Lv Haicai , Guo Wentao , Liu Fang , Niu Fenglei
Published: November 2024

To study the natural circulation capabilities of Liquid Lead Bismuth (LBE), natural circulation loops have been designed and constructed in several countries for natural circulation characterization of LBE. These loops aim to characterize LBE’ natural circulation properties. Seeking a more profound comprehension of LBE behavior during natural circulation, this research leverages a Computational Fluid Dynamics (CFD) simulation program. This program theoretically models the circulation loop, while a one-dimensional code is employed to verify the model’s accuracy. The findings reveal that the discrepancies between the two models are minimal, thereby affirming their trustworthiness. Utilizing the validated model, a deepening of the influence of loop geometric parameter, heating power, and the configuration of the pooled solid-state oxygen control device on the natural circulation flow rate within the lead-bismuth loop. Furthermore, the CFD model is instrumental in examining the thermal and hydraulic attributes of the loop under distinct operational scenarios, namely transient startup and power jump conditions. In essence, this paper integrates two robust models to explore the thermal and hydraulic nuances of the liquid lead bismuth recirculation loop. Through this exploration, the optimal geometric parameters are established and an assessment of the viability of the pool-type solid-state oxygen control device is also performed.
Effect of Geometric Parameters on Cooling Performance of The Concrete Shielding for Swimming Pool-Type Low-Temperature Heating Reactor
Authors: Fan Guangming , Hao Shuai , Li Zongkun , Li Jiaming , Guan Shicong
Published: November 2024

Swimming Pool-type Low-Temperature Heating Reactor (SPLTHR) is a reactor developed to meet the needs of clean heating in northern China. But the core inlet/outlet temperature of SPLTHR is only about 68/98 °C. In order to ensure the heating capacity of SPLTHR in winter, it is necessary to increase the outlet temperature of the reactor. Increasing the outlet temperature of SPLTHR will certainly lead to the rise of the pool water temperature, which will affect the bearing capacity and service life of the concrete around the reactor. It is necessary to reduce the concrete temperature around the reactor. Based on the design of the flowing water layer scheme, the numerical simulation of a full-scale Swimming Pool-type Low-Temperature Heating Reactor was carried out to analyze the effects of geometric parameters on the cooling performance of the concrete. The results show the the radial position of water injection, the number of water injection, the depth of connecting holes, the thickness of the water layer have a significant impact on the cooling performance. The maximum radial temperature difference between the pool water and the pool wall can reach 12.06°C.
Study on the Calculation of Interphase Drag Force Based on Liquid Metal-Gas Two-Phase Flow
Authors: Zhang Hong , Ou Yangyong , Wang Di , Zhu Longxiang , Wan Lingfeng , Sun Wan , Pan Liangming
Published: November 2024

The steam generator (SG) is the core equipment for heat transfer of lead-based fast reactor. Due to the large temperature difference and pressure difference between both sides of the SG heat transfer tube bundle, the steam generator tube rupture accident (SGTR) may occur. After the SGTR accident, the high-pressure water on the secondary side enters the primary side and is heated to generate a large number of bubbles, which would threaten the nuclear safety. Therefore, it is essential to predict the void fraction in liquid metal-gas two-phase flow. An accurate calculation of interphase drag force is the key to predicting the void fraction. To achieve this, this paper experimentally studies the characteristics of liquid Lead Bismuth Eutectic (LBE)-gas two-phase flow in a vertical pipe with diameter of 9 cm and length of 2 m, calculates the drag force by drift velocity method with different model parameters and adopts one dimensional two-fluid model with the four order Runge-Kutta method for predicting the void fraction. This paper obtains the predicted void fraction under different gas phase fluxes and discusses the influence of key model parameters on the calculation.
Numerical Study on the Operational Behavior of Catalytic Components With Spherical Packed Bed Structures
Authors: Man Tianming , Guo Zehua , Ding Ming , Liang Wenkai
Published: November 2024

In order to mitigate the potential risk of hydrogen explosion in severe accidents, Passive Autocatalytic Recombiners (PARs) have become widely employed in the containment atmosphere of PWR nuclear power plants in recent years. Catalytic components with spherical packing forms, characterized by a larger catalytic specific surface area, show significant potential to enhance the hydrogen removal capability of PARs. Three typical structured packing forms are considered: simple cubic (SC), body center cubic (BCC) and face center cubic (FCC) arrangements of packed catalyst spheres. The numerical models of catalytic components were constructed using the CFD software STAR-CCM+. The elementary reaction mechanism is employed as the reaction kinetics model to describe the reaction process. When the inlet hydrogen concentration is 4.0 vol.% and the inlet velocity is 0.3 m/s, the results indicate that hydrogen removal efficiencies for the SC, BCC, and FCC arrangements of packed catalyst spheres are 81.47%, 98.89%, and 99.87%, respectively. Notably, the FCC catalytic component exhibits the highest hydrogen removal capability. However, the most complex structure and densest arrangement of the FCC catalytic component also introduce the highest flow resistance coefficient, reaching 43.04.
Review of Transient Analysis and Accident Simulation Codes for Nuclear Reactor
Authors: Qiu Xinze , Liu Tao , Tong Jiejuan , Zhang Liguo , Wu Yuzeng
Published: November 2024

In the background of global carbon emission reduction, nuclear energy plays an important role in mitigating global warming. Safety is the cornerstone of nuclear energy development. Nuclear reactor transient analysis and accident simulation codes can be used for safety assessment to evaluate the design and operation of reactors, thus reducing the occurrence of nuclear accidents. Efficient nuclear reactor transient analysis and accident simulation codes can also help operators make decisions in emergencies. In this paper, we review the past nuclear reactor transient analysis and accident simulation codes for LWR and some advanced reactors. We summarize and classify different codes. We find that the system analysis codes mainly include thermo-hydraulic models, reactor dynamic models, and control models. Overall the code size is huge and the computational cost is high. Local transient analysis codes are less computationally intensive and more accurate for, for example, the simulation of transients introduced by reactivity. The integrated transient analysis code combines the advantages of each code, but it requires data transfer between different codes, resulting in high computational costs. The accident process simulation code focuses on the presentation of the accident evolution trend. This paper also describes the future direction of transient analysis and accident simulation codes. Both the transient analysis code and the accident simulation code will need to become more specialized and more generalized in the future. Accurate physical models and computational efficiency are important for transient analysis codes. Speediness and lightweight are important for accident simulation codes. Finally, the role of AI in code model design is presented.
Effect of Hydrogen Plant Structure on Hydrogen Diffusion
Authors: Shucheng Zhang , Lyu Xuefeng , Wang Lin , Li Xichen , Yu Yu , Zhao Houjian , Wang Shengfei , Guo Zhangpeng
Published: November 2024

Nuclear energy is a primary energy source that is low in carbon and highly efficient. The hydrogen production process coupled with nuclear power produces clean, low-carbon, and highly efficient hydrogen. However, hydrogen leakage accidents may occur during the hydrogen production process, and the structural design of hydrogen production plants significantly impacts the risk of hydrogen combustion after an accident. This paper introduces two models for electrolytic water hydrogen plants: the flat roof vent model and the pointed roof vent model. The impact of plant structure on hydrogen behavior and hydrogen explosion risk inside the plant was investigated using computational fluid dynamics software Fluent. The results indicate that in the event of a hydrogen leakage accident, the hydrogen concentration inside the plant with a pointed roof and ventilation outlets is relatively lower. Especially, the accumulation of hydrogen in the area above the leak point on the roof of the plant is significantly improved compared to a flat roof structure with ventilation outlets. In the same duration, the pointed roof structure facilitates faster accumulation of hydrogen at the top of the plant and expedites its expulsion through the ventilation outlets, so the overall residual hydrogen quantity is reduced. It is highly recommended that the design and construction of hydrogen production plants take into account the possibility of hydrogen leakage accidents. In order to mitigate such risks, plant structures should feature spire roofs along with ventilation openings.
CFD Simulations of Helium Flow Through Partial Heating Pebble Bed
Authors: Li Jiajie , Ji Yu , Sun Jun
Published: November 2024

The modular high temperature gas-cooled reactor (HTGR) is an advanced technology of inherent safety, high electricity generation efficiency, and potential process heat applications in the chemical industry. The packed pebble bed in HTGR consists of a large number of randomly distributed pebbles and is cooled down by helium flow. The reactor thermal hydraulic design and safety analysis requires in-depth understanding of helium flow and heat transfer, and reliable correlations in the pebble bed. Previous research mostly focuses on the helium flow in non-heating pebble bed (in hot tests before criticality) or full heating pebble bed (full of fuel pebbles releasing fission or decay heat). Understandings on partial heating pebble bed are insufficient. Current paper uses the Computational Fluid Dynamics (CFD) to study the helium flow in partial heating pebble bed with mixed fuel and graphite pebbles, which is typical in the initial core and running-in phase of pebble bed HTGR. Various flow conditions in the pebble bed were simulated, and the pressure drops were compared with those predicted by Kerntechnischer Ausschuss (KTA) equation. The Reynolds number and porosity used in KTA equation were discussed in calculating the pressure drops. It is found that the pressure drop of uniform heating model and nonuniform heating model are basically the same. The pressure drop predicted by KTA equation is larger than the calculated result. As for the using of KTA equation, the effect of Reynolds number is not significant while the porosity is of great importance. This work lays a certain foundation on the development of a new pressure drop correlation, which could support the design and commissioning of the pebble bed HTGR.
Numerical Prediction of Flow and Heat Transfer in Supercritical Fluids
Authors: Alasif Abdullah , Pucciarelli Andrea , Siddiqui Osman , Shams Afaque
Published: November 2024

Enhancing the thermal physical properties of coolants can be achieved by multiple means. One of the easiest ways is to compress and heat the fluid above the critical point to the supercritical state. Supercritical fluids have the potential to be ideal coolants. For that reason, many Gen-IV reactor designs are considering using supercritical fluids as coolants and working fluids. However, surpassing the critical point introduces complexities in predicting the thermodynamic behavior of these fluids. This has crippled the widespread implementation of supercritical fluids in thermal-sensitive applications like nuclear reactors. Extensive experimental studies have been conducted to understand the complex heat transfer behavior of supercritical fluids and more importantly to unravel the mystery behind the deterioration and enhancement phenomena. These studies serve as a reference to assess the available Computational Fluid Dynamics (CFD) based turbulence models. The current work aims to assess the newly developed advanced Reynolds-Averaged Navier-Stokes (RANS) models for predicting the heat transfer of supercritical fluids. In particular, the prediction capabilities of different advanced Algebraic Heat Flux Models, namely: AHFM-UniPi and AHFM-SC, are tested for two flow conditions within a circular tube. In addition, a sensitivity analysis of the AHFM-SC coefficients is presented, and a new set of coefficients is proposed.
OpenFOAM-Based Benchmarking Simulations of Xe Gas Behavior in Molten Lead for Advanced Fast Reactors
Authors: Hanic Emir , Wang Xiang
Published: November 2024

In the process of researching fast reactors that use molten lead as a cooling fluid, we realized that none of the projects deals with the problem of the creation of radioactive gas from the core. In further research, we realized that no one deals with this problem either experimentally or using computer simulations, and this paper attempts to deal with this issue by defining a benchmark study that could be used in the future to evaluate any potential simulations that would simulate Xe as a gas in the molten lead. We used OpenFOAM in which we defined radioactive decay in the transport equation and incorporated it into the new solver reactorFoam. We tested our methodology with known cases with water and air. We also investigated results from the commercial software COMSOL Multiphysics, and we have compared these results with results obtained with OpenFOAM. After the methodology has been established and confirmed, the change of parameters was introduced to Xe and molten lead. In this paper, we present the first results of this type, and we believe that due to the difficulty of performing such an experiment, this simulation is a potential candidate for a benchmark study between Xe and molten lead.
Research on Depressurization Strategy of AP1000 Automatic Depressurization System Under DVI Pipeline Break Accident
Authors: Lv Dufeng , Shang Zhengrun , Meng Zhaoming , Sun Zhongning
Published: November 2024

In this paper, based on the system analysis program, the AP1000 nuclear power plant is taken as the research object, and the AP1000 system analysis model is built. Taking the DVI pipeline break accident as the starting event, the depressurization strategy of the ADS is studied. The results show that after the DVI pipeline break accident, the primary loop’s pressure can be gradually reduced to the input pressure of the in-containment refueling water storage tank (IRWST) through the multi-stage pressure relief of ADS, so that the three types passive water tanks can be orderly injected to ensure that the cladding does not overheat. There is a flow suppression area in the safety injection process of the core make-up water tank (CMT). In the DVI pipeline break accident, the first three levels of ADS have little effect on the input of high-pressure and medium-pressure safety injection, and the input time of ACC in different ADS pipeline fault conditions is only 2 seconds apart. In the DVI pipeline break accident, under the ADS pipeline fault conditions simulated in this paper, the low-pressure safety injection can be successfully put into operation to ensure the safety of the core. The multi-stage design of ADS can gradually increase the inlet pressure of the sparger. Opening ADS-2 or ADS-3 with larger size at the initial time can keep the sparger in a stable critical jet state for a short time. However, in the DVI pipeline break accident, the sparger cannot maintain a stable critical jet for a long time due to the large break of the pipeline and the strong pressure relief ability. Therefore, at the beginning of nuclear power design, it is necessary to consider that the jet flow pattern of the sparger in the DVI pipeline break accident will be in the chugging and condensation oscillation area, accompanied by large dynamic load and noise, which poses a potential threat to the integrity of the sparger and the safety of the nuclear power system.
Analysis of Hydrogen Transport Operation Under Multiple Drive Sources in Containment
Authors: Liu Xinyan , Gao Li , Hu Zongwen , Sun Zhongning , Bian Haozhi , Ye Zixiang
Published: November 2024

As a kind of clean energy, nuclear energy has great development potential and application prospects, but we must take into account the possible harm to the environment and society caused by radioactive material leakage. Therefore, the integrity of the containment vessel, which is the third and final barrier against the release of radioactive materials, must be effectively guaranteed. When a water loss accident occurs in a nuclear reactor and causes a serious accident, a large amount of steam will be released from the pipe breach, and hydrogen will be released into the large space of the containment vessel due to the zirconium-water reaction. When the concentration of hydrogen reaches a certain level, explosion may occur, seriously damaging the integrity of the containment. In this paper, a full-scale model of the containment vessel and its key equipment is established by means of numerical simulation to analyze the transport and diffusion behavior of hydrogen under the action of multiple driving sources such as heat exchangers and breaks. The results show that the sedimentation effect caused by the condensation of water vapor on the surface of the heat exchanger will affect the transport direction of the surrounding hydrogen, and the existence of compartments will also hinder the flow of hydrogen. The accumulation of hydrogen in large space can be effectively alleviated by optimizing the design of compartments.
Research on the Film Thickness Characteristics of Jet Condenser Nozzle
Authors: Chu Weiyu , Huang Chonghai , Xiao Qi , Meng Zhaoming , Fu Yilin
Published: November 2024

Owing to the expansive applicability of jet condensers across diverse sectors, juxtaposed with the paucity of scholarly investigation into the morphological characteristics of the condenser liquid film, the present study zeroes in on the determinants of film thickness attributable to jet condenser nozzles. In pursuit of this objective, we engineered and actualized a visualization experimental apparatus tailored for the meticulous examination of the film-forming propensities of such nozzles. Subsequent experimental endeavors focused on the dynamics of nozzle-induced spray film genesis. Leveraging bespoke measurement techniques, this inquiry delved into the influence of varied parameters including flow rate and nozzle angle, on the film thickness engendered by the nozzles, thereby elucidating discernible trends in the phenotypical attributes of nozzle film thickness. Findings from this research suggest a proportional relationship between film thickness and flow rate, and an inverse correlation between nozzle diameter and film thickness. These insights furnish empirical underpinnings for the refinement of nozzle configurations in jet condensers, thereby contributing to the enhancement of their operational efficacy.
Investigation of the Heat Transfer Properties of Natural Circulation Flow in the Irradiation Channel of the HFETR
Authors: Jin Shuai , Sun Sheng , Yang Wenhua , Si Junping , Xu Yixing , Lei Jin
Published: November 2024

The High Flux Engineering Test Reactor (HFETR) is a pressurized water reactor with a 125 MW operating capacity that uses beryllium as a reflector and light water as a moderator and coolant. Moreover, it may perform research on how various reactor fuels and materials function under irradiation inside the reactor. Using a separator tube, HFETR has created a number of stagnant water channels to keep the cooling water inside the channels separate from the reactor’s cold removal water. This work describes the use of COMSOL Multiphysics software to construct a flow heat transfer calculation model for the irradiation test of instrumentation materials in the HFETR static water channel. Based on measurements of the cooling water temperature in the channel under different power steps, the accuracy of the calculation model was confirmed. This paper studies the temperature and velocity distribution of the cooling water in the channel under different power steps, and calculates the average convection heat transfer coefficient at the device wall.

The results show that the HFETR’s static water channel has a good natural circulation capability, which improves the safety of irradiation test. This also establishes the foundation for the appropriate irradiation tests in the static water channel and the thermal analysis and computation.
Current Results of BME NTI in the ECC-SMART Project: Different Analysis on the SCW-SMR
Authors: Varju Tamás , Kiss Attila , Antók Csenge , Babcsány Boglárka , Boros Ildikó , Mészáros Péter , Bertesina Zeno , Czifrus Szabolcs
Published: November 2024

BME, as a consortium member, is participating in the ECC-SMART project, which has been awarded an EU/EURATOM/H2020 grant in 2020, and is researching the development of a supercritical water-cooled small modular reactor concept called SCW-SMR. This paper summarizes the research work carried out so far at BME within the framework of the project. Firstly, several international benchmarks were modelled by the APROS system code and the Ansys CFX CFD software package in order to demonstrate their applicability for the thermal hydraulics analysis of SCW. Later, a complete reactor vessel and a simplified primary circuit model were built with APROS, and a full-length fuel assembly model was developed in Ansys CFX. All the seven heat-up stages of the SCW-SMR pre-conceptual design were modelled and analyzed with these tools. The Serpent 2 Monte Carlo code and the in-house SPNDYN finite element code were used for reactor physics benchmarking, for various stand-alone reactor physics calculations and in loosely coupled (with APROS and Ansys CFX) and tightly coupled (with APROS) code systems in order to design and optimize the fuel and the reactor core of the SCW-SMR concept. Many successful calculation series (e.g. the so-called SCWR-FQT benchmark) have been performed with these codes and code systems.
Risk-Informed Defense-in-Depth Strategy for Nuclear Power Plant
Authors: Liu Qianwen , Wu Yuxiang , Wang Yuhan
Published: November 2024

As one of the basic safety design theory, Defense-In-Depth (DID) concept has been proven effective for ensuring the safety of nuclear reactors by safety operation of hundreds of units for decades. The classic defense-in-depth concept containing five defense layers is built on mainstream light-water reactor type. And after Fukushima nuclear accident, improvements of the DID concept was proposed by IAEA and WENRA, including adding sublevel for additional Design Extension Condition (DEC), and emphasizing the independence between defense levels. However, there are several problems during DID practical application; and current DID concept based on deterministic method has limitations for innovative design or advanced technology application, such as non-light water reactor designs, because current DID strategy evolves from LWR type and earlier deterministic hypothesis. How to develop a more comprehensive, holistic, and technology-inclusive DID strategy is explored in this study. A risk-informed (RI) method is adopted to improve the DID strategy by combining the traditional DID concept with a risk-informed approach. Considering the gradual development of risk assessment methods and tools and acceptance process of risk insight for nuclear industry, a progressive risk-informed DID strategy is proposed in this article, including three RI solutions with different optimized elements, which are implemented through an integrated risk-informed DID design process balancing between deterministic and risk perspectives, and allocating resources to weakness that might be overlooked.
Numerical Research on Heat Transfer Characteristics of Helium-Xenon Mixture in 19-Rod Bundle
Authors: Xu Lisha , Qiu Zicheng , Xu Jianjun
Published: November 2024

Helium-xenon mixture is one of the coolant options for high-temperature gas-cooled reactors (HTGR). As an important factor affecting reactor core design, the convective heat transfer performance of the coolant has always been the focus of research on HTGR. The helium-xenon mixture is a low Prandtl number gas, and its thermodynamic properties deviate from the ideal gas equation. Based on the experimental parameters in the literature, this study used FLUENT to construct a CFD model, and introduced the polynomial function obtained by fitting the physical property data calculated by the Tournier formula into the model. By comparing the wall temperature, fluid center temperature and average temperature in the experimental results and simulation results, the effects and errors of different turbulence models and particle transport models on the convective heat transfer results of helium-xenon mixture are studied. It is found that the standard k-ϵ model can well simulate the flow and heat transfer process of helium-xenon mixture in the tube, and the maximum local temperature difference does not exceed 5%. Based on this research, a helium-xenon mixture flow heat transfer model of 19-rod bundle is established. This study observe the convective heat transfer process of helium-xenon mixture under extreme conditions and reference helium-xenon reactor design parameters, and explore the changes in the average Nu number, local Nu number, and frictional resistance coefficient with temperature, pressure, and flow field. It provides a reference for the design and optimization of future HTGR.
OECD-NEA Expert Group on Reactor Systems Multi-Physics: Development of Verification and Validation Guidelines for Multi-Physics Simulations
Authors: Valentine Timothy E. , Ivanov Evgeny , Ivanov Kostadin , Petruzzi Alessandro , Avramova Maria , Hursin Mathieu , Buss Oliver
Published: November 2024

Over the past decades, national and multi-national efforts have been established to develop high-fidelity, multi-physics, and multi-scale modeling and simulation (M&S) tools for design and safety challenges for existing and advanced reactors. These tools are being developed and utilized to provide predictive capabilities that enhance safety, improve designs, and accelerate innovation. An expert group on reactor systems multi-physics (EGMUP) was established in 2014 by the OECD Nuclear Energy Agency (NEA) to address specific challenges with validation of such M&S tools with the intent to provide member countries of the NEA with consensus guidance and recommendations and to demonstrate the application of the validation principles. The EGMUP supervises several multi-physics benchmark exercises to develop and demonstrate best practices for validation. This paper will describe the overall mission of the group, provide a description of several of the validation exercises relevant to existing reactor technologies, and summarize key recommendations for validation approaches.
A Kind of Shielding Design and Verification Test of Radioactive Source Transport Container
Authors: Chen Lei , Wang ZhiPeng , Jiao LiMing , Zhuang DaJie , Sun Qian , Wang ChangWu
Published: November 2024

A transport container for radioactive sources has been meticulously designed to facilitate the global transfer of radioactive materials from storage facilities to end-users. In compliance with the stipulations outlined in GB 11806 and the IAEA SSR-6, the radiation dose of the container was meticulously calculated under both normal transport conditions and accident scenarios. Shielding tests were conducted both at the manufacturing site and post-manufacturing. The results of these tests indicated that the surface dose rate was 289 μSv/h prior to testing, which increased to 299.34 μSv/h after testing. Additionally, the radiation level at a distance of 1 meter from the package was measured at 31.59 μ Sv/h. Both the computational and experimental outcomes confirm that the shielding design of the transport container adheres to the relevant standards set forth by GB 11806 and SSR-6.
REVIEWS OF ACOUSTICAL PATENTS
Authors: Fulop Sean A.
Published: November 2024

The purpose of these acoustical patent reviews is to provide enough information for a Journal reader to decide whether to seek more information from the patent itself. Any opinions expressed here are those of the reviewers as individuals and are not legal opinions. Patents are available via the Internet at the USPTO website.
Office soundscape assessment: A model of acoustic environment perception in open-plan offices
Authors: West Benjamin , Deuchars Angus , Ali-MacLachlan Islah
Published: November 2024

The soundscape approach has been utilised in recent research for residential environments, but applications to indoor working environments remain limited. This study investigated the suitability of affective quality evaluations of open-plan office acoustics. Perceptual assessments were completed by occupants of eight open-plan office floorplates, using affective dimensions outlined in PD ISO/TS 12913-3:2019 and indoor soundscape studies. Participant demographic, contextual, work-related quality, and psychological well-being data were collected. Workstation noise levels were measured following BS ISO 22955:2021. Principal component analysis identified Pleasantness, Eventfulness, and Emptiness as the main perceptual dimensions, cumulatively explaining 56% of the total variance. Results indicate the suitability of the PD ISO/TS 12913-3:2019 two-dimensional model for open-plan offices. The perceived presence of human sounds negatively correlated with ISO Pleasantness yet positively correlated with ISO Eventfulness. Participant gender and aural diversity mediated ISO Pleasantness. Psychological well-being and work-related quality correlated positively with soundscape pleasantness. No correlations were found between level-based acoustical indicators—equivalent continuous sound level (LAeq), level exceeded 10% (LA10) and 90% (LA90) of the time, level variability (LA10–LA90) and Liveliness—and perceptual scores. A conceptual representation of open-plan office soundscape perception is presented, offering a framework for evaluating and designing workspaces to enhance well-being.