bluuebunny/crossref_metadata_2025_split huggingface dataset:
Heat Transfer to Supercritical Water Flowing in Short Vertical Bare Tubes
Authors: Kavalci Mehmet , Wspanialy Mark , Heyns Laura , Cornelius Marcus , Pioro Igor 
Published: November 2024

Abstract
It is well known that water-cooled nuclear-power reactors / Nuclear Power Plants (NPPs) equipped with light- and heavy-water-cooled reactors have relatively low thermal efficiencies within the range of 26–38% compared to those of thermal power plants – combined-cycle up to 62.5% and supercritical-pressure Rankine cycle up to 55%. Therefore, increasing thermal efficiency of next generation NPPs / nuclear reactors is a very important task. For water-cooled reactors / NPPs the conventional approach for increasing thermal efficiency is to move to supercritical pressures, as it was done more than 60 years ago in the thermal-power industry.
Currently, together with large-capacity SuperCritical Water-cooled Reactors (SCWRs) SC-Pressure Small Modular Reactors (SCP SMRs) are also being developed. Therefore, solid knowledge of heat-transfer specifics at supercritical pressures is required. Usually, hydraulic-equivalent diameters of modern bundle flow geometries are within 5–10 mm and heated lengths are up to 4 m for SCWRs and can be shorter for SCP SMRs.
Also, first SCWRs as test, experimental, or pilot reactors will be with relatively small installed capacities within 100–300 MWel. Due to that it is important to investigate supercritical heat transfer in small diameters tubes as a first and preliminary step instead of bundle geometries and bundle simulators. Usually, heat transfer in bare tubes is considered as the conservative approach compared to bundle tests, because heat transfer in bundle geometries will be higher compared to that in bare tubes due to additional turbulization from various appendages inside sub-channels.
The paper presents experimental data obtained in relatively short vertical bare tubes mainly with upward flow of SCW. The experiments have been performed by Dr. V. Razumovskiy1 at the National Technical University of Ukraine. The data were obtained at the pressure of 23.5 MPa, mass fluxes of 2193 kg/m2s (main set of data) and 509 and 248 kg/m2s, and heat fluxes within 104–2660 kW/m2.
The experimental data were compared to a well-known subcritical-pressure heat-transfer correlation developed by Dittus and Boelter. The heat-transfer-coefficient data were predicted quite reasonably with this correlation within the normal heat-transfer regime, but were significantly higher those within the deteriorated heat-transfer regime and within the pseudocritical region.
Analytical Study of Melt Spreading in Shallow Pool Observed in PULiMS-E10 Test Using MSPREAD Implemented With Molten Jet Quench Model
Authors: Kikuchi Wataru , Hotta Akitoshi , Ito Koetsu , Yugo Hiroaki , Shimizu Mamoru 
Published: November 2024

Abstract
The subject of this study is a numerical simulation of the melt spreading in a shallow pool using MSPREAD, a two-dimensional high-resolution melt spreading analysis code. The identified influential factors in under-water melt spreading are the heat transfer between the melt and the overlying water, melt-coolant interaction (MCI) caused by water confinement at the time of molten jet impingement, molten jet breakup into melt slugs, disperse of melt slugs and cooling by water recirculating inside the partially solidified debris bed. In particular, the above-mentioned MCI and the subsequent chain of phenomena were named “the molten jet quench” by the authors, and a numerical model was proposed and implemented in MSPREAD. In this model, the diameter of spherical melt slugs and the radius of the floor region where melt continues to be slugs can be given as the user inputs. The three sensitivity analysis cases were performed with varying these two inputs together with no quench cases based on the PULiMS-E10 test conducted by the Royal Institute of Technology. Comparison of the time histories of the spreading area, pool water temperature, 2D spreading shapes and post-test debris cross section indicated that the molten jet quench can explain short spreading distances, thicker solidified debris, and even higher pool water temperatures observed in PULiMS-E10.
Research on Typical Fault Prediction Model of the Reactor Coolant Pump Based on Multi-Information Fusion of GRU
Authors: Huaiming Cui , Chengxiao Kuang 
Published: November 2024

Abstract
According to the sensor type, measuring point distribution and acquisition frequency of the Reactor Coolant Pump (RCP) in the nuclear power plant, the typical failure mode of the key vulnerable parts of RCP is investigated and sorted out. Then, the simulation test bench of bearing and shaft seal components is set up respectively to simulate the operation of bearing and shaft seal components in several typical faults, and the multi-channel fault sample data is collected. In order to solve the problem of poor prediction accuracy and weak generalization ability of traditional neural network. Based on the above experimental data, a multi-information fusion prediction model of key signal of main pump based on gated circulation unit GRU is proposed in this paper. After testing the test set data, compared with the traditional deep learning algorithm, it is shown that the prediction model designed in this paper has the best prediction effect.
Researchers Understanding of the Possible Contribution of International Atomic Energy’s to the Launch of a National Nuclear Power Programme
Authors: Ala Ayodeji , Oluwafolakemi Ala 
Published: November 2024

Abstract
The continuous rise in global energy demand is a result of factors such as increasing globalization, increased access to technology, increasing world population, and rising standard of living. This study investigates the level of understanding of the role of the International Atomic Energy Agency among researchers involved in nuclear power programmes. The level of knowledge on the launching of a nuclear power programme is significantly related to the existence of a nuclear power plant in a country while the understanding of the IAEA milestone approach correlates with the knowledge of other nuclear technology applications in the country. These results show that research reactors and the wide application of nuclear technology in a country can contribute to understanding the critical issues associated with nuclear power programme development.
SAMPO-P Test Facility and Numerical Pretest
Authors: Guo Pengya , Yu Peng , Quan Fengyang , Li Wei , Pei Jie , Yuan Yidan , Yu Jiyang , Ma Weimin 
Published: November 2024

Abstract
The in-vessel retention (IVR) strategy was developed to retain the corium within the reactor pressure vessel (RPV) to prevent the release of radioactive products during severe accidents. However, the RASPLAV study reveals the potential formation of a two-layer melt pool pattern, wherein the heat flux in the top metal layer significantly exceeds the average heat flux density, posing a considerable threat to the integrity of the pressure vessel and the success of the IVR strategy. Previous studies on stratified melt pools have primarily concentrated on the natural convection behavior within the melt pool using scaled facilities. To better understand the thermal behavior of the two-layer melt pool heat transfer, this study constructed a 1:1 ratio-sized 2D stratified melt pool experiment facility ‘SAMPO-P’ based on the HPR1000 RPV lower head. The numerical pre-test simulations are conducted using in-house code to help determine the experimental conditions and test matrix. The simulants of the oxide layer and metal layer were Chlorine salt and Al-Mg alloy. The results indicate that as the heating power increases from 180 kW to 255 kW, the temperature and heat flux increase while the thickness of the bottom crust decreases with no significant heat flux peak in the metal layer observed. The first experiment is scheduled for this year. The experimental results are expected to provide essential insights into understanding the flow and heat transfer characteristics and further improve the numerical model in future research.
A Liquid Metal Subchannel Code Used in Ocean Conditions and Study the Heat Transfer Characteristics in Rod Bundle of Reactor Core
Authors: Yin Yuanyuan , Han Bin , Zhu Xiaoliang , Qi Siwei , Liu Shenghui , Xing Tianyang , Yang Bao-Wen , Liu Aiguo 
Published: November 2024

Abstract
In this study, a thermal hydraulic ocean version subchannel analysis code is developed for liquid metal reactor core analysis based on the theoretical model of ocean conditions. In the developing process, the ocean conditions are divided into heaving, rolling and pitching conditions. During the development of the program, it is found that the research on the characteristics of flow resistance and heat transfer mechanism under ocean conditions is not mature, the theoretical models are not perfect, and there is no widely recognized empirical relationship under ocean conditions. The original empirical model is still used in the first version of the program. Furthermore, the verification and validation of the subchannel program is carried out by comparing the results of experimental data. The effects of geometric parameters such as rod Pitch to Diameter ratio P/D, Rod to Wall gap (RTW) are studied. In addition, the boundary conditions of inlet mass flow on the flow and heat transfer under ocean conditions are investigated. The effect of rolling motion on different coolants including liquid meal and water are discussed under ocean conditions. The newly developed subchannel analysis code can well complete the thermal and hydraulic calculation of liquid metal cooled fuel assemblies under ocean conditions, and can provide support for the design and thermal and hydraulic analysis of reactor in ocean environment.
Study on the PSA Application in Emergency Operating Procedure Development
Authors: Ma Chao , Feng Churan , Du Jinyan , Yang Jian 
Published: November 2024

Abstract
With the development of PSA technology, more and more PSA technology is being applied in various design fields during the design process of nuclear power plants. The emergency operating procedures for nuclear power plants are an important part of accident response and processing. From previous PSA analysis results, it can also be seen that emergency operating procedures play an important role in ensuring the safety of nuclear power plants. The emergency operating procedures are in line with the design of nuclear power plants. The power plant operating conditions considered in the design of nuclear power plants need to be reflected in the emergency operating procedures, and the measures for responding to nuclear power plant accidents need to be considered in the accident. The emergency operating procedures should cover the main event sequences of the power plant. In addition to relying on past engineering practices, PSA can provide a more systematic analysis of event sequences, and provide support for the development of emergency operating procedures based on the likelihood of accidents and event sequences occurring, as well as the overall risk of the power plant. This article combines the technical characteristics of PSA and applies PSA technology to the process of preparing emergency operating procedures for nuclear power plants. It provides a general method and beneficial insights for PSA to support the preparation of emergency operating procedures, making them more systematic, complete, and effective. PSA provides a systematic analysis from potential accidents to the final handling of accidents, and supports emergency operating procedures in the following two aspects: the conditions considered by emergency operating procedures (including design basic conditions and design extension conditions); Is the emergency operating procedure sufficient and effective in guiding the actions of important operators, especially in operator actions under extension operating conditions.
Neutronics/Thermal-Hydraulics Coupling Simulation Using JAMPAN in a Single BWR Fuel Assembly
Authors: Kamiya Tomohiro , Nagatake Taku , Ono Ayako , Tada Kenichi , Kondo Ryoichi , Nagaya Yasunobu , Yoshida Hiroyuki 
Published: November 2024

Abstract
We have aimed to realize high-fidelity neutronics/thermal-hydraulics coupling simulation to provide simulation results that can be used as validation data for reactor analysis codes. We have developed a multi-physics platform, JAMPAN, to conduct neutronics/thermal-hydraulics coupling simulation by connecting independent codes. It is required to reduce empirical correlations as far as possible to perform high-fidelity neutronics/thermal-hydraulics coupling simulation. Hence, a continuous energy Monte Carlo code MVP is adopted as a neutronic analysis code, and a detailed and phenomenological numerical simulation code JUPITER is adopted as a thermal-hydraulics analysis code. Our simulation target is a single BWR fuel assembly. Hence, the simulation results need to reproduce the varying heat generation owing to the varying void fraction. Therefore, we carried out MVP/JUPITER coupling simulation in a 8 × 8 single fuel assembly of BWR and confirmed that the void fraction and heat generation distribution are reasonable qualitatively. Furthermore, it is required to clarify the effect of the parameters of the coupling simulation on the results, and the time interval is one of the coupling parameters to improve the reliability of the simulations. We carried out MVP/JUPITER coupling simulation in a 2 × 2 fuel pin system using the multi-physics platform JAMPAN to investigate the effect of the time interval on the results. A 2 × 2 fuel pin system, which is the smaller unit of the actual single fuel assembly for a BWR, is adopted as a simulation target to effectively investigate the effect. It is found that the effect of the time interval on the simulation results is small.
Status of the Light-Water, Small, Modular Reactors Research at Oregon State University
Authors: Gutowska Izabela , Howard Trevor Kent , Wu Qiao , Woods Brian G. 
Published: November 2024

Abstract
This paper provides a comprehensive overview of the ongoing research endeavors conducted at Oregon State University School of Nuclear Science and Engineering (OSU NSE) in the field of Light-Water, Small, Modular Reactors (LW-SMR). Included in this paper is a thorough exposition of the integral testing activities carried out at the Multi-Application, Small, Light-Water, Reactor (MASLWR) facility, including both the methodologies employed and the outcomes obtained. The test facility can operate at pressure of 11.4 MPa and a temperature of 590 K. It serves as an integral test facility, with all components designed at a 1/3 scale height and a 1/254.7 volume scale in reference to the NuScale module. The facility operates in compliance with the NQA-1 program and has been instrumental in supporting NuScale’s licensing process and the validation of TH (thermal-hydraulic) codes. This paper further expounds on the present operational status of the facility and delineates upcoming plans. Beyond the experimental capabilities, additional attention is directed towards an explication of the research capabilities intrinsic to the Nuscale Simulator control room, housed within the OSU Energy Exploration (E2) Center. Finally, the paper delves into an exploration of ongoing research concerning the Dynamical System Scaling (DSS) of Small Modular Reactors (SMRs).
This paper serves as a consolidated resource offering insight into the multifaceted aspects of LW-SMR research conducted at Oregon State University, encompassing integral testing, facility status, control room research capabilities, and the DSS endeavors related to SMRs.
Effects of Mixed Inert Gases on Design of Centrifugal Compressors for Space Nuclear Reactor
Authors: Xu Risheng , Ma Wenkui , Ye Ping , Zhao Gang , Yang Xiaoyong 
Published: November 2024

Abstract
The aerospace industry considers nuclear reactor as the prime candidate to extend mission durations and enhance the energy supply for spacecraft. Among the power conversion systems for space reactors, the direct closed Brayton cycle stands out for its compactness, lightness, safety, and high power density, making it exceptionally well-suited to these applications. Centrifugal compressor is a key component of the Brayton cycle, whose performance is pivotal in the system efficiency. Researches on compressors utilizing inert gases has predominantly concentrated on helium-xenon mixtures, leaving other inert gas combinations less explored. Moreover, the influence of inert gas mixture proportions on compressor design has not been adequately addressed in the literature. To fill this research gap, this paper established models of the inert gas mixtures properties and design of centrifugal compressors, and investigated the effects of inert gas mixtures, including helium-argon and helium-neon mixtures, on performance and geometric parameters of the centrifugal compressor. The results indicated that helium-argon or helium-neon mixtures working fluid could significantly reduce the number of compressor stages, thus potentially minimizing the compressor’s volume and mass compared to pure helium working fluid. The increase in the molar mass of the gas mixture decreased the rotational speed and maximum circumferential velocity of the compressor and increase the mass flow rate and the Mach number at the exit of the final-stage impeller. This study provides a reference for the selection of working fluids in space reactor Brayton cycles and the design of centrifugal compressors.
A Coordinated Control Strategy for NHR200-II Under Load Rejection Condition
Authors: Huang Canxing , Xiong Huasheng , Dong Zhe , Li Boyuan , Zhou Shuqiao 
Published: November 2024

Abstract
In recent years, the safety and stability of power plant operations have attracted more and more attention. As a clean and safe energy, nuclear power is up-and-coming in China. However, external interruption is hard to avoid and usually causes negative consequences. The load rejection condition may occur in nuclear power plants due to environmental influences, external power grid shock, or failure. This paper mainly discussed the load rejection process of NHR200-II, one of the steam generator outlet valves closed unexpectedly and the reactor load changed from 100% to 50%. The load rejection condition was simulated in MATLAB SIMULINK and analyzed in detail. A control strategy for this abnormal condition was proposed and a cascade control over the steam pressure control loop was introduced. This control strategy was validated by conducting simulation experiments. By comparing the responding curves of critical parameters before and after the control is applied, the result showed that the severe consequence could be avoided after applying the control strategy and the stable operation was realized.
Effect of Structure-Soil-Structure Interaction on the Seismic Response of the Nuclear Building of Small Modular Reactor
Authors: Wang Dongyang , Sun Xiaoying , Zhang Chaoqi 
Published: November 2024

Abstract
The seismic Category I structures within the nuclear island of the Small Modular Reactor (SMR) are densely clustered and connected through the underlying foundation soil. This study discusses the phenomenon of structure-soil-structure interaction (SSSI), where the soil medium transmits seismic responses from one structure to its adjacent structures, employing the sub-structuring method. Seismic response analyses are conducted for the nuclear island structures of SMR, with and without considering the SSSI effect. The impact of SSSI on the seismic responses of the nuclear buildings is investigated by comparing their in-structure response spectra (ISRS). The results indicate that SSSI has a minimal influence on the dynamic response for structures with large volumes, as no significant changes are observed in the acceleration response spectra shape, dominant frequency position, peak acceleration, and zero period acceleration value of the ISRS. However, an evident impact is observed on the dynamic response for the standby diesel generator building due to the SSSI effect. Although the response spectra shape remains virtually unchanged, a slight decrease in main frequency is detected after considering the SSSI effect, along with variations in the spectral peak acceleration value and zero period acceleration value to a certain extent.
Vigorously Develop Nuclear Power to Meet the Challenge of the Rapid Growth of Electricity Consumption and Carbon Emission in the Information Industry
Authors: Li Yingnan , Wu Fang , Tang Te , Liu Weili 
Published: November 2024

Abstract
The information industry is developing rapidly, but the rapid growth of the electricity demand in information industry has not been paid enough attention. According to our research of this paper, the electricity demand in information industry will exceed 10,000 TWh (10 trillion kilowatt-hour) in 2030. As a result, on the one hand, the speed increase of global installed capacity should be 60% higher than the current forecast, in order to meet the electricity demand of information industry. Additionally, due to the acceleration of carbon emission across the information industry, it will be more difficult to achieve the goal of 1.5°C temperature rise limit by 2030. On the other hand, in the historical process of green adjustment and low-carbon transition of the global energy structure, the electricity cost in the information industry will also increase significantly. Namely, with the rapid development of the information industry, the information industry itself, global power supply, carbon emission and electricity price balance are facing more severe challenges. According to the study in this paper, it is believed that vigorously developing nuclear power is the most economical and realistic way to effectively deal with these challenges.
Experimental Study of Bubble Migration Characteristics in MCCI Molten Pool
Authors: Yang Guorui , Zhang Jing , Wu Yingwei , He Yanan , Qiu Suizheng , Su Guanghui , Tian Wenxi 
Published: November 2024

Abstract
When the molten core-concrete interaction (MCCI) occurs in a reactor, the high-temperature melt may melt through the concrete bottom plate and form a molten pool. At the same time, due to the heat generated by decay heat and chemical reactions, the concrete will decompose and produce a large amount of water vapor and carbon dioxide and other gases. After entering the molten pool, these gases will migrate within it and have a stirring effect on the molten pool. This study focuses on the phenomenon of gas migration in MCCI molten pool, and establishes an experimental platform for studying the migration characteristics of gas in MCCI. A 25% glycerol aqueous solution was used as a substitute fluid for the melt, and helium gas was used as a substitute gas to study the migration characteristics of gas in the molten pool. The experimental platform has 49 ventilation holes, which can simulate the migration of bubble groups in the molten pool, and each ventilation hole can adjust the apparent gas velocity separately. The experiment mainly measured the migration velocity, equivalent diameter, and aspect ratio of bubbles. By adjusting the gas flow rate, the changes in these three parameters were studied to obtain the migration characteristics of bubbles. Through this study, it was found that with the increase of apparent gas velocity, the equivalent diameter of bubbles increases, the aspect ratio of bubbles decreases, and the final stable velocity of bubble migration increases. Through this study, the variation patterns of bubble size, shape, and velocity under different gas flow rates were obtained, which can provide support for systematic calculation programs to calculate chemical reactions, fission product release, and molten pool stratification in MCCI processes.