bluuebunny/crossref_metadata_2025_split huggingface dataset:
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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

Abstract
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.