bluuebunny/crossref_metadata_2025_split huggingface dataset:
Study on Reactor Cavity Cooling System of HTR-PM in DLOFC Accident With TIN-CAVCO Code and SPECTRA Code
Authors: Xu Xinsheng , Ye Yiyang , Zheng Yanhua , Marek M. Stempniewicz 
Published: November 2024

Abstract
High Temperature Gas-cooled Reactor (HTGR) is recognized by the international nuclear power industry as a safe type of reactor. The modular pebble-bed high temperature gas-cooled reactor uses reactor modules with relatively small thermal power between 200–600MWt, utilizing the high-temperature resistance that can be achieved by coated fuel particles. Institute of Nuclear and New Energy Technology (INET) of Tsinghua University has developed a commercial-scale 200 MWe High Temperature gas-cooled Reactor Pebble-bed Module project (HTR-PM). The decay heat can be removed by natural mechanism, such as thermal radiation, heat conductivity, natural convection, thereby guaranteeing the safety of reactor under accident and eliminating the possibility of core melting. A passive Reactor Cavity Cooling System (RCCS) has been designed for HTR-PM to ensure the safety of the Reactor Pressure Vessel (RPV) and reactor cavity under normal operation and accident conditions. RCCS transfers heat from pressure vessels to the atmosphere through thermal radiation, heat conduction, and natural convection, and the system utilizes passive mechanism with high safety and reliability. The RCCS of HTR-PM has been designed as three independent sets. According to the preliminary design, two sets work can guarantee the safety of the PRV and reactor cavity.
In this paper, the working performance of RCCS under depressurized loss of forced cooling (DLOFC) accident has been analyzed by INET with TIN-CAVCO code, and also by NRG with SPECTRA code. The analysis results of these two codes are in good agreement, proving the applicability of both codes in analysis of the HTGR as well as its RCCS. During the accidents, the maximal fuel temperature keeps below the limitation of 1620°C. The results also prove the excellent performance of the RCCS. The RCCS meets the design requirements, that is the normal operation of two sets can effectively ensure the safety of the pressure vessel and the reactor cavity during the accident. Even if only one set works normally, the decay heat still can be carried out effectively and the RPV temperature will not exceed the limitation of 425 °C.
A Preliminary Study on the Transport of Critical Radionuclide Polonium in the Marine Environment After Source Term Release
Authors: Zeng Junyang , Wang Yuqing , Zhang Minghao , Ni Muyi 
Published: November 2024

Abstract
Regional Ocean Modeling System (ROMS) was employed to model the transport of radionuclide Po-210 in the ocean over a one-week period, simulating the scenario of a hypothetical nuclear accident. The simulation incorporated considerations for the radioactive decay of Po-210, various release modes, and the combined impact of fifteen tidal constituents. The initial release concentration of Po-210 was set at 10000 unit/m3, with the release point situated near a coastal nuclear power plant in Guangdong Province, China.
The concentration distribution analysis revealed that, driven by ocean currents, Po-210 initially moved towards the northwest coast from the release point. Subsequently, it followed a southwesterly trajectory along the coast, impacting the entire coastal region of western Guangdong Province. Five days postrelease, the influence of Po-210 extended to the South China Sea.
Analyzing the concentration distribution and results from three coastal concentration detection stations revealed consistently high concentration values in coastal waters, indicating a significant impact of polonium on the coastal environment. Notably, high Po-210 concentrations in the waters surrounding Port X (the estuary) in Guangdong Province suggested a potential impact on inland rivers, directly affecting nearby residents. Cross-sectional analysis unveiled that Po-210 exhibited rapid vertical mixing, a phenomenon closely associated with tidal influences and water depth. Comparing simulation results with and without tidal effects demonstrates that that the rising tides exerted an inhibitory effect on Po-210 transport, while ebbing tides had a promoting effect.
Steam-Water Two-Phase Flow Testing Under Nominal Pressure and Temperature Conditions for Natural Circulation Performance of Chimney in BWRX-300 SMR
Authors: Furuichi Hajime , Povolny Antonin , Katono Kenichi , Yasuda Kenichi , Kito Kazuaki , Heck Charles 
Published: November 2024

Abstract
The BWRX-300 is a water-cooled, natural circulation, small modular reactor developed by GE Hitachi Nuclear Energy in cooperation with Hitachi-GE Nuclear Energy. BWRX-300 is equipped with passive safety systems and leverages the design and licensing of USNRC-certified ESBWR. The BWRX-300 uses a tall chimney above the core to provide natural circulation head to drive sufficient core flow and ensure fuel is operated within its thermal limits. Thus, calculated void fraction in the chimney is a key quantity for estimation of the natural circulation core flowrate. BWR transient analyses are performed with Transient Reactor Analysis Code, GE version (TRACG) which has been qualified for calculating thermal limits in BWRs. This extent of qualification includes the evaluation of core flow rate and evaluation of void fraction in chimney or upper plenum of BWRs. To support extending the TRACG qualification of void fraction evaluation to the large-diameter chimney of BWRX-300, additional test data were obtained with steam-water two-phase flow test and the void fraction modeling and calculation in TRACG were validated. The test was conducted in Hitachi Utility Steam Test LEading facility (HUSTLE) using a chimney test section, which was down-scaled to around 0.5 m in diameter. Distributions of the void fraction and the gas-liquid interfacial velocities in the chimney were measured by differential pressure gauges and the Wire-mesh Sensor (WS) for nominal BWR operating conditions of 7 MPa and 290 °C and flow conditions scaled from conditions of BWR-300 chimney at a range of reactor power levels. The average void fraction in the chimney estimated by the TRACG model and compared with the HUSTLE data. Consequently, the qualification of the void fraction estimation of the TRACG model was extended to the large-diameter chimney of BWRX-300.
Simulation Study on the Transient Operating Characteristics of Natural Circulation Coupled With PCM Energy Storage Module for the Passive Containment Cooling System
Authors: Huang Yufan , Wang Han , Chen Weixiong , Yang Xiaohu , Li Shaodan , Yan Junjie 
Published: November 2024

Abstract
To achieve a further reduction in pressure peak during accidents, phase change materials (PCM) are utilized in conjunction with natural circulation in the passive containment cooling system (PCCS). The dynamic simulation model of PCCS has been developed using Apros software. This article focuses on a comparative analysis of cooling capacity, considering various thicknesses and types of PCM modules. The results indicate that employing a 4 mm PCM module decreases the pressure peak in the containment from 1.85 MPa to 1.68 MPa. Additionally, the impact of thickness on the maximum power of the heat exchanger is constrained by the lower thermal conductivity of the PCM. Another type of PCM with the same latent heat capacity but lower thermal conductivity maintains pressure peak suppression performance. However, it results in a diminished pressure drop performance during the continuous condensation of steam. As a consequence of increased accumulation of non-condensable gases at the bottom, the heat transfer for steam condensation at the corresponding position is impeded.
Neutronics Study and Conceptual Design for Small Sodium-Cooled-Fast-Reactor With Annular Fuel
Authors: Ye Bin , Ala Ayodeji A. 
Published: November 2024

Abstract
This paper presents a conceptual design for a small Sodium-Cooled-Fast-Reactor core with annular fuel through the analysis of neutronic character, various power strategies and control rod control system. The design work of the conceptual core is based on BN800. Take the universality of fuel assembly into account, the hexagonal fuel assembly cladding of the conceptual core is consistent with BN800. The inner and outer diameter of the annular fuel are 4.0mm and 11.8mm separately, and the mixed oxide (MOX) fuel enrichment from the inside to the outside of the core are 19.5%, 22.1%, and 24.7%, respectively. The number of fuel assemblies in the inner/middle/outer region of the core is 109/102/156. The finite multiplication factor (keff), Neutron flux (Φ), neutron energy spectrum, effective delayed neutron fraction (βeff), sodium void reactivity effect (αv), and fuel burnup performance parameters of the reactor are all studied, furthermore, the control rod performance with different absorbing materials (boron carbide (B4C), hafnium diboride (HfB2), and hafnium hydride (HfH1.62)) are also compared and evaluated. This research work provides a reference for the optimization design of dual-cooled annular fuel for small sodium-cooled fast reactor.
Study on Heat Transfer Behavior of a Rectangular Particle Bed With Volumetric Heating
Authors: Junlang Wen , Yuto Kamada , Kosei Yokoyama , Tatsuya Matsumoto , Wei Liu , Koji Morita , Yuya Imaizumi , Hirotaka Tagami , Kenichi Matsuba , Kenji Kamiyama 
Published: November 2024

Abstract
During core disruptive accidents (CDA) in sodium-cooled fast reactors, the core debris bed could be formed due to the interaction between the molten fuel and coolant. At the late stage of CDA, the heat removal of decay heat will be achieved mainly by the natural circulation of the molten stainless steel to the surrounding environment. To investigate the coolability of fuel debris bed, a rectangular experimental system was built in which the particle bed was volumetrically heated via direct current heating. The experimental apparatus consists of a Nd-Fe-B magnetic particle bed immersed in water and a water pool above it, which simulate disrupted solid fuel and molten steel, respectively. Magnetism ensures the good contact between particles; therefore, the stable volumetric heating is truly achieved, which is similar to the characteristic of fuel debris with internal decay heating. Computer code simulations with the reactor safety analysis code SIMMER-IV were performed to help understand the heat transfer characteristics and to validate the applicability of the newly embedded momentum exchange function (MXF) models. The time variation of average temperatures of particle bed and water pool, and their difference were evaluated to compare with the experimental results. The comparison results show the most applicable MXF model under current experimental conditions and well reproduce the influence of water pool height in the experiments.
Inventory Control Strategy and Dynamic Modelling of a 1 MWe HeXe Brayton Cycle Reactor System
Authors: Li Xinyu , Guan Chaoran , Chai Xiang , Liu Xiaojing 
Published: November 2024

Abstract
The small mobile nuclear power system using the HeXe Brayton cycle as its energy conversion system has promising application potential. Among the research, control strategies get a lot of attention, which makes variable operation of the system possible. Among various control strategies, inventory control is widely used in the load-following process as it enables the system to remain high efficiency during load regulation. However, compared to other control strategies, it has a slower response time, which needs to be improved. In this study, a one-dimensional dynamic model of a simple HeXe close Brayton cycle reactor power conversion system, as well as its main equipment, has been proposed. Focusing on inventory control, the effects of initial tank volume and pressure on load range that the control strategy could meet were investigated. The open loop simulation of coolant mass in the loop was carried out. After that, to overcome its disadvantage, inventory control has been modified for rapid load regulation by combining with valves, so that the optimized inventory control method could meet the rapid load change process while maintaining high efficiency and safe operation of the system.
A Study on the Burnup Effect of Silver Self-Powered Neutron Detector
Authors: Sang Yaodong , Deng Bangjie , Liu Shiyu , Zhang Qingmin 
Published: November 2024

Abstract
Self-Powered Neutron Detector (SPND) has been widely used in reactor core for in-core neutron flux monitoring. As a typical delayed type of SPND, the burnup behavior of Ag SPND is more complicated than Rh SPND and VSPND, because Ag is composed of two natural isotopes 107Ag (51.839%) and 109Ag (48.161%) with distinct neutron capture cross-sections. This will lead to change in neutron sensitivity of Ag SPND during operation, which in turn affects the monitoring of neutron flux. In this work, the burnup effect of Ag SPND utilized in CE-type assembly was studied by using SPNDSignal toolkit. By solving the burnup equations, we calculated the composition change of the emitter material over 10.5 operation years. Ag SPNDs with different burnup rates were modeled in SPNDSignal, and the components contributing to neutron sensitivity were calculated and analyzed. After 10.5 years of use, 7.84% of 107Ag and 21.08% of 109Ag are burned up, and correspondingly it leads to neutron sensitivity decrement by 16.68%; however, the sensitivity to external gamma is negatively small and remains almost unchanged during the whole operation period. On the other hand, signals’ timing analysis shows the rising time of the total signal current to a step neutron flux increases from 115 s to 125 s. All of these suggest that the neutron sensitivity and delay compensation coefficients need to be recalibrated during the use of Ag SPND.
Preliminary Investigation on Burst Behaviors of Zr-Based Cladding Under Simulated LOCA Conditions
Authors: Duan Zhengang , Yuan Bo , Wen Qinglong , Chen Kang 
Published: November 2024

Abstract
The Fukushima accident exposed the shortcomings of zirconium-based fuel cladding in withstanding severe loss of coolant accidents (LOCA), thereby accelerating the development of accident-tolerant fuel (ATF). Over the past decade, extensive technical exploration has led to chromium (Cr) coated zirconium cladding emerging as the primary technological route slated for commercial application soon. In current study, the burst behaviors of zirconium-based claddings were systematically investigated under simulated LOCA conditions. The initial internal pressures ranged from 2–10MPa, and the heating rates were 14 °C/s and 28 °C/s. Although research on Cr-coated cladding is still ongoing, preliminary findings from the zirconium-based cladding analysis indicate that at a consistent heating rate, the burst time decreases with increasing internal pressure, and the burst temperature decreases accordingly. Conversely, at a constant initial internal pressure, the burst time decreases as the heating rate increases, but the burst temperature elevates.
Effect of Heat Treatment on Microstructure Characteristics of Hot Worked Zr-2.5Nb Alloy Tubes
Authors: Yin Jianming , Gao Bo , Luan Baifeng , Zhang Yan 
Published: November 2024

Abstract
Characterization and analysis techniques including XRD, FEG-SEM with attached EBSD detector, TEM with attached EDS detector were employed to investigate the microstructure characteristics of a Zr-2.5Nb alloy tube. The samples as received from the hot worked tube were treated under 550°C/5.5h annealing, and 850°C/1h solution heat treatment followed by 550°C/5.5h aging (SAHT) respectively. Hardness test was also performed for predicting of the strength changes. Results show that there was residual β-Zr existed in the hot worked tube, and after annealing at 550°C, decomposition of the residual β-Zr in the hot worked tube happened and the low angle grain boundaries reduced. There is no obvious texture memory effect after solution at 850°C flowed by water quenching, however the effect can be well observed after aging at 550°C. Nb-rich second phase particles appeared during aging process, with their locations near the boundaries inside primary a grains. Hardness of the hot worked tube didn’t show obvious change after annealing, while after solid solution and aging treatment (SAHT), the hardness increased significantly both in LD and TD directions.
Machine Learning Approach to Analyzing Data in Japanese BWR Chemistry Database
Authors: Yamazaki Gaku 
Published: November 2024

Abstract
Understanding the relationship between the dose rate of the primary loop recirculation (PLR) piping and chemistry parameters is the beginning of developing the techniques of reducing the radiation exposure in boiling water reactors. In this study, a regression model was built using a machine learning method, gradient boosting, to identify the relationship between the dose rate of the PLR piping and several chemistry parameters.
Categorical Boosting, EXtreme Gradient Boosting, and Light Gradient Boosting Machine were used to build a regression model, with the dose rate of the PLR piping as the target variable and chemistry parameters as the explanatory variables. The coefficient of determination and root mean squared error across these three models are nearly identical, demonstrating the similar accuracy of different gradient boosting methods.
SHapley Additive exPlanations values were used to interpret the internal structure of these models and rank the importance of parameters associated with the dose rate of the PLR piping. All three models highlighted the same top-tier parameters, indicating that comparing regression models could be beneficial for enhancing the robustness of parameter extraction. Only within certain ranges was a strong relationship found between the dose rate of the PLR piping and these chemistry parameters, implying non-linear relationships and optimal ranges corresponding to lower values of the dose rate of the PLR piping.
Research on Fluid-Structure Coupling of Spent Fuel Storage and Transport Basket Based on Immersed Boundary Element
Authors: Zhang Feifan , Liu Yu , Lu Daogang , Xie Fei 
Published: November 2024

Abstract
Nuclear power plants are intricate systems housing a multitude of equipment crucial for their functioning. Among these components, the spent fuel storage and transport basket plays a pivotal role in securely containing highly radioactive spent fuel. Particularly under seismic conditions, the spent fuel storage and transport basket encounters challenges associated with fluid dynamics. Fluid permeates its interior, and external fluid dynamically envelops it, introducing complex fluid-structure coupling effects. Traditional fluid-structure coupling models, characterized by closed structures, tend to adopt overly conservative approaches. Additionally, transient Computational Fluid Dynamics (CFD) simulations, relying on intricate modeling, pose computational challenges due to their intensity. In response to these challenges, this research introduces an innovative approach by leveraging the immersed boundary method. The research not only proposes a new methodology for studying fluid-structure coupling phenomena but also utilizes the OpenFOAM software for numerical simulations. The study provides comprehensive insights into the variation of lift force, contributing to a deeper understanding of the fluid-structure interactions. To validate the proposed approach, the numerical results undergo a rigorous comparative analysis with both experimental data and previously modeled outcomes. In conclusion, through a meticulous and comprehensive numerical investigation, the findings not only offer a foundation for structural dynamic analysis but also hold implications for enhancing the seismic resilience of crucial components within nuclear power facilities.
Accident Progression Analysis of an Offshore Floating Nuclear Power Plant Under Heaving Motion
Authors: Ui Atsushi , Ikeda Yutaka , Arai Takahiro 
Published: November 2024

Abstract
Offshore Floating Nuclear Power (OFNP), in which a proven commercial light water reactor (LWR) is mounted on a floating structure such as an oil drilling rig has been proposed by the Massachusetts Institute of Technology. Based on this concept, the Japanese industry is conducting research and development on the feasibility of an OFNP with a boiling water reactor (BWR) on a floating structure and operating 30 km off the coast of Japan. As part of this study, safety analysis was conducted to confirm whether the heaving oscillation affects accident progression in design extension conditions during the oscillation of OFNP, which is assumed to occur in stormy weather and other high-wave conditions. Prior to the analysis, the terms including gravitational acceleration in the equations of motion of the reactor system analysis code RELAP5/MOD3.3 were modified to handle acceleration during oscillation. In addition, experiments corresponding to the oscillation were conducted under the same conditions of approximately 290°C and 7.2 MPa as the BWR core, and the experimental results were used to validate the code. After confirming that the code was able to reproduce the void fraction variations of the experiment, uncertainty analysis of accident progression, such as TQUV and ATWS, including various heaving conditions, was performed to analyze the effects of the oscillations.
A Brief Introduction to Several Important Design Optimizations of Passive Nuclear Power Plants in China
Authors: Ma Baisong , Miao Zhengqiang , Ma Yuanhua 
Published: November 2024

Abstract
Passive nuclear power technology was developed by Westinghouse in the United States and introduced to China around 2005. Currently, four passive nuclear power plants have been operating safely and stably in China for about 20 reactor years. However, some design defects have also been exposed. Therefore, a series of design optimizations have been carried out to improve the reliability and ease of operation and maintenance in passive nuclear power plants. This paper briefly introduces the several important design optimizations implemented and planed in Chinese passive nuclear power plants. We hope it can provide useful reference for other countries in the construction and operation of the passive nuclear power plants. (1) A gate has been added in refueling cavity. It creates conditions for equipment maintenance during refueling operations, which require the water level in the Reactor Pool to be lower than the level of flange of the Reactor Vessel. (2) A spare Variable Frequency Drive (VFD) has been added to solve the problem of unreliable long-term operation of the VFD for Reactor Coolant Pump (RCP). A 50Hz RCP has been selected for the plants under construction. (3) The anchor point of the support for Main Stream Pipelines has been adjusted from the Auxiliary Building in Nuclear Island to the adjacent wall of the First Bay so as to control Main Control Room (MCR) noise. The noise source has been also reduced by reducing the main stream flow rate. (4) A Spare Glove Box has been added, which is standby with the Grab Sample Panel. According to the characteristics of the sampling object, the sampling pipeline shall be reasonably separated to reduce or avoid the mixing of the sampling liquid, so as to improve the sampling representativeness.