bluuebunny/crossref_metadata_2025_split huggingface dataset:
Numerical Analysis of Thermal Stratification for Upper Plenum in Pool-Type Sodium Fast Reactor
Authors: Xing Mingdi , Cheng Daoxi , Ma Xiaoyao , Zhai Weiming , Zhou Ping , Yu Song , Li Ruizhi , Gao Weilong 
Published: November 2024

Abstract
During the scram transient of China demonstration sodium fast reactor, thermal stratification phenomenon is observed in the upper plenum. The paper simulates the thermal stratification which occurs when hotter sodium from the core outlet enters the upper plenum along with colder sodium from the DHX entering the bottom of upper plenum using ANSYS commercial code. The results illustrate three transient stages of thermal stratification: In the first stage, hotter sodium accumulate in the top of the plenum while interface continues descending; simultaneously colder sodium from the DHX enter into plenum forming another thermal stratification phenomenon with rising interface until both interfaces meet forming a stable interface near the core outlet; In the second stage, as the flow rate and temperature at the core outlet gradually decrease, the hotter sodium cannot fully reach the top of the upper plenum, and another interface is formed at the top. In the third stage, the thermal stratification interface at the top gradually disappears, and the final thermal stratification interface is formed near the core outlet. Furthermore, it is also concluded that the interface descend rate decreases with the increase of Richardson number.
Optimization Design and Application Analysis of HLW Glass Container Under High Level Liquid Waste Separation Strategy
Authors: Duo Ma , Meng Wei , Jing-Yang Wang 
Published: November 2024

Abstract
This paper analyzes the “high heat release” logistics source term under the strategy of spent fuel reprocessing and high-level liquid waste separation. It calculates the annual production of high-level waste glass in a reprocessing plant with an annual processing capacity of 800 tons under different limiting conditions. It is found that according to the current design scheme, the thermal power of a single glass solidified cylinder is the determining factor for the production of high-level waste glass. Based on this, this article optimizes the design of strengthened thermal conductivity for high-level waste glass containers at La Hague post-treatment plant in France, and conducts thermal calculation analysis under its application conditions using ANSYS software to optimize the design scheme. The main findings of this study are: 1) Under the high level radioactive separation strategy, by strengthening the thermal conductivity of high-level waste glass containers, the theoretical maximum production of high-level waste glass can be reduced by 47.4% (reaching a limit of about precious metal content); 2) Adding a thermal barrier in the high-level waste glass container can reduce the maximum temperature of the high-level waste glass by 79 °C under certain conditions; 3) Through preliminary quantitative analysis, the optimized high-level waste glass container design scheme can reduce the production of high-level waste glass by about 30%.
Comparative Study of Cluster Analysis Methods in Dynamic Safety Analysis of Nuclear Power Plants
Authors: Suliman Mohamedelmogtabh Omer Elfadni , Wang He , Xiong Binfu 
Published: November 2024

Abstract
In nuclear power plant safety analysis, employing dynamic safety analysis methods such as Dynamic Event Trees (DET) to analyze individual initiating events often generates numerous scenarios, posing challenges in extracting useful information. This study utilizes clustering analysis methods to categorize these accident scenarios and extract their main features, thereby simplifying the analysis process and enhancing efficiency. Different clustering analysis methods can yield varying results, affecting the ability to reflect accident processes and extract key features. To ensure that clustering analysis better reflects accident processes and extracts the main features of accidents, this study compares and investigates two clustering analysis methods: one that considers only extreme parameter values of accidents and another that considers parameters throughout the entire accident process. In the method focusing solely on extreme parameter values, key parameters are extracted as clustering analysis parameters from each accident scenario. In contrast, the method considering parameters throughout the entire accident process treats accident scenario parameters as time-series data and applies Dynamic Time Warping (DTW) algorithm for time-series clustering.
Using the Steam Generator Tube Rupture (SGTR) accident as an example, this study applies both methods to analyze SGTR accidents and obtains corresponding results. The characteristics and differences between the two methods are analyzed.
Investigating the Effect of Supports on Reinforced Concrete Slab Subjected to Oblique Missile Impact
Authors: Kang Zuoyi , Okuda Yukihiko , Nishida Akemi , Tsubota Haruji , Itoh Masaharu , Li Yinsheng 
Published: November 2024

Abstract
As an outer protective engineering structure for nuclear power plant (NPP), the reinforced concrete (RC) containment wall at NPP should be evaluated for safety requirement related to external threats. Therefore, much attention has been paid to investigate the impact resistance ability of the RC slab structure subjected to missile impact. In these studies, most of them focused on local damage of RC slab structures struck by a rigid missile along the normal direction of the impact surface, while only few studies have focused on oblique impact. So, we have conducted a series of impact tests under different impact conditions covering impact angles, stiffnesses of missiles and so on. The objective is to figure out the different impact behaviors of the RC slab structures. We also intended to put forward a numerical method based on the test results, and to validate the proposed method through comparisons between experimental and numerical results.
In the test results, the reaction forces of RC slab were measured by using the load cells installed on the four corners of RC slab’s back surface. In the oblique impact, a special support structure was manufactured to fix the RC slab. It is speculated that the different supporting structure may influence the test results of reaction force. Thus, this work concentrates on the effect of the stiffnesses of the supports for oblique impact on the reaction forces of RC slab. Until now, static loading tests were carried out to confirm the stiffnesses of the components of supporting parts. The findings obtained from the comparison between the numerical results using the experimental values as spring constants in the finite element (FE) model and the reaction force measurement results will be reported.
A Study on Analysis of Actual Data in “Fugen” Decommissioning
Authors: Kodua Yuya , Nakamura Yasuyuki , Iguchi Yukihiro , Yanagihara Satoshi 
Published: November 2024

Abstract
Decommissioning Engineering Center (“Fugen”), has been proceeding with decommissioning work since receiving approval for the decommissioning plan in 2008, and is currently in the second step of dismantling the reactor peripheral equipment.
’Fugen” has a structure other than the reactor that is similar to a BWR, so it will be reflected in future commercial reactors.
In decommissioning, project management is important to select strategies and optimize processes, waste, safety, costs, etc. In Japan, it is expected that decommissioning work for nuclear facilities will become more serious in the future, and the knowledge and project management data obtained from the decommissioning work carried out at Fugen to date will be useful for planning.
Since the dismantling and removal of the main turbine equipment and the peripheral equipment in preparation for the reactors have been completed so far, the actual data has been organized and analyzed and compiled into unit work coefficients.
The project management data (Man-powers, amount of waste, etc.) collected during the decommissioning work carried out at “Fugen” is organized and calculated as a unit work coefficient (Man-powers required per unit weight (person/hour/ton)).
In summary, we analyzed the differences in workability due to differences in the working environment.
Improvement of the Thermal-Hydraulic Feedback Model in Bamboo-Core From Single-Channel to Sub-Channel
Authors: Zhang Hengrui , Li Yunzhao , Wang Sicheng , Li Yisong , Liang Yiling 
Published: November 2024

Abstract
The thermal-hydraulic feedback is critical for PWR-core analysis. The real flow distribution at the core inlet shows some non-uniformity and varies with operating conditions. Most core analysis codes assume that the core inlet flow distribution is uniform. To minimize the power distribution variations due to different core inlet flow distributions, the sub-channel model is employed to replace the single-channel model in the thermal-hydraulic feedback module of Bamboo-Core, which is a full-core reactor core analysis code for the Pressurized Water Reactor (PWR). The new sub-channel thermal-hydraulic feedback model is validated by comparing its calculation results with the measured data from the first three operational cycles of M310. When both uniform and non-uniform core inlet flow distribution were used to calculate the CNP650. The maximum bias between two core inlet flow distribution is 8.2%. Subsequently, the bias of 3D power distribution due to two different flow distributions was calculated. For single-channel thermal feedback, the maximum bias is 0.598%, while for sub-channel feedback, the maximum is 0.205%. The Changes in core inlet flow distribution have less impact on the results with sub-channel thermal feedback than single-channel thermal feedback. Therefore, the sub-channel thermal feedback is more suitable for the case of approximate processing of coolant flow distribution at the core inlet than the single-channel thermal feedback.
Research on Energy-Saving Operation Optimization of Circulating Water System Under CHP of Nuclear Energy Comprehensive Utilization
Authors: Wu Fang , Tang Te , Liu Weili , Li Yingnan , Wu Xinzhuang 
Published: November 2024

Abstract
Nuclear energy heating is the most elementary and the most essential mode of comprehensive utilization of nuclear energy. Based on the principle of variable operating conditions of the turbine, the principle of energy balance and mass balance of the thermal system, and the method of calculating condenser pressure, we investigated the impacts of extracting steam for heat supply on the distribution of steam and water parameters and thermal characteristics of the turbine, and determined the heat balance diagram of the turbine after changing the rated condensing condition to heat supply and the power-backpressure characteristics of AP1000 nuclear turbine. Based on the thermal characteristics of AP1000 nuclear turbine under rated heat supply condition, we investigated the effect of changing the number of operating fixed-speed circulating water pumps on the unit economics. Based on the optimal vacuum principle, we analyzed the frequency conversion renovation and operation optimization of AP1000 nuclear turbine circulating water pumps under heat supply condition, and identified the optimal renovation scheme and economic benefits.
A Comparative Analysis of CFD Approaches to Model Wire-Wrapped Fuel Bundle and Experimental Validation
Authors: Halim Ossama , Pucciarelli Andrea , Forgione Nicola 
Published: November 2024

Abstract
A wire-wrapped fuel bundle of 19 rods using Lead-Bismuth Eutectic (LBE) as a coolant was chosen for this CFD study and the predictions are compared to the NACIE-UP benchmark experimental data. The k-ω SST turbulence model is used to investigate the heat transfer of the LBE together with the adoption of a constant turbulent Prandtl number of 1.5. The thermo-dynamic properties of the LBE are modelled as temperature-dependent polynomials. Three different approaches have been adopted in this work to model the FPS, namely, Bare, Detailed, and the Porous Rod Bundle model. A model-to-model comparison is performed to understand the benefits, limitations, and accuracy of using porous models for representing wrapped wire. Furthermore, the integration of NACIE-UP benchmark experimental data into the comparative analysis adds an additional layer to validate the modelling approach.
Coupled Thermal-Hydraulics and Neutron Transport Calculations of Small Modular Reactor Using Serpent, OpenFOAM and SUBCHANFLOW Codes
Authors: Kořínek Tomáš , Zavorka Jiri , Lovecky Martin , Skoda Radek 
Published: November 2024

Abstract
Nowadays, several Small Modular Reactors are under development and high-fidelity neutron transport calculations with thermal-hydraulic feedback are a powerful tool for design and optimization of such reactors. The present work focuses on analyses of heavy water small modular reactor using coupled CFD, Monte Carlo and subchannel code. The coupling of Monte Carlo and subchannel codes was based on Picard iterations with power relaxation using stochastic approach. CFD simulations served for a generation of inflow conditions for subchannel code. The coupled pin-by-pin subchannel analyses was compared against fuel assembly level subchannel analyses. Additionally, a case with asymmetrical inflow conditions was analyses using coupled codes to reveal influence of inflow uncertainty on criticality and maximum fuel temperatures.
Comparison of coupled calculations on pin-by-pin level and fuel assembly level showed that a lower resolution led to underprediction of maximum fuel temperatures by 108 °C and maximum cladding temperature by 9 °C. Further, the low resolution led to the underprediction of multiplication factor by 410 pcm.
The asymmetrical inflow case led to a slightly higher coolant, cladding, and fuel temperatures in order of few degrees due to a lower flow rate in half of the core. Further, the asymmetric condition did not influence the reactivity as the multiplication factor changed insignificantly.
Constrained Nonlinear Model Predictive Control for High Temperature Gas-Cooled Reactor in Load Following
Authors: Zhu Yunlong , Dong Zhe , Huang Xiaojin 
Published: November 2024

Abstract
As the climate issue has become the focus of global attention, new energy with less carbon emissions have attracted the attention of the world, especially nuclear energy. High temperature gas-cooled reactor is an inherently safe reactor designed by the Institute of Nuclear and New Energy Technology of Tsinghua University and has been put into commercial operation. The nuclear steam supply system (NSSS) is the key system of the modular high temperature gas-cooled reactor for generating superheated steam, including a reactor core and a once-through steam generator (OTSG). In the process of nuclear power plant operation, power-level control is the key technology to ensure the NSSS to operate in the desired state. By adjusting and controlling the control rod speed, the rotation-rate of the main helium circulator and the valve-opening, the neutron flux, helium temperature and water supply flow rate are adjusted, and then the power-level maintenance, maneuvering and load tracking are realized. Considering the strong nonlinearity of NSSS, it is necessary to develop nonlinear power-level control methods. In this paper, model predictive control (MPC) is proposed to realize load following in the NSSS of high temperature gas-cooled reactor. MPC has the advantage of obtaining control input by iterating and solving optimization problems, which has a good application scenario for power-level control of nuclear power plants. The research in this paper has an enlightening significance for the application of MPC to the power-level control of high temperature gas-cooled reactor. In the future engineering application, facing the actual situation of multi-objective and multi-constraint conditions, the MPC iterative optimization idea can be used to obtain the control input that can make the nuclear power plant stable and meet the operation requirements.
Study on Single-Phase Flow and Heat Transfer Characteristics of Vertical Narrow Rectangular Channel Under Rolling Motion with Nonuniform Heating
Authors: Di Zhu , Rulei Sun , Yang Li , Sichao Tan , Ruifeng Tian 
Published: November 2024

Abstract
Plate type fuel elements have wide application potential in the integrated reactors due to their high power-to-volume ratio, low core temperature, high burnup and compact structure arrangement. The flow channels between adjacent fuel element plates are small, showing typical narrow rectangular structures with large aspect ratio. Due to the influence of fuel self-shielding effect, irradiation in the reactor and component arrangement, the plate fuel element has obvious uneven heat release in the transverse direction. As a result, the thermal hydraulic characteristics in the flow channel may be different from those in the uniform heating and conventional channels. In this paper, a typical vertical narrow rectangular channel is selected as the research object, and the channel section is 65mm × 2.23mm. Based on the computational fluid dynamics method, the mathematical and physical model of flow and heat transfer is established, and refined numerical calculations are carried out. The research parameters are that the inlet temperature is 38°C, the outlet gauge pressure is 0MPa, the heat flux is 15–240kW/m2, and the Reynolds number is 1000–14000. The research results show that the effect of transverse non-uniform heating on the heat transfer characteristics is greater than that on the flow characteristics. And with the decrease of Reynolds number and the increase of heat flux, the influence of non-uniform heating is more significant. Subsequently, the flow heat transfer analysis under the rolling and pitching conditions of the ocean is carried out to analyze the influence of non-uniform heating. It is found that Nusselt number and friction coefficient are macroscopic consistent with vertical state under forced circulation ocean conditions. However, under the same rolling period and Angle, the secondary flow intensity in the rolling state is greater than that in the pitching state. In the rolling state, different nonuniform heating forms have obvious influence on the secondary flow, in which the linear distribution form has the highest secondary flow intensity.
Method Research on the Model Optimization of Whole-Core Fuel-Assembly Bowing Based on 3D Variational Algorithm and the Measurement Values
Authors: Guo Lin , Zhang Kai , Wan Chenghui , Wu Hongchun 
Published: November 2024

Abstract
The phenomenon of fuel-assembly bowing would inevitably occur during operation, but the fuel-assembly bowing cannot be measured directly. Therefore, the challenge lies in obtaining the actual fuel-assembly bowing, which is a prerequisite for ensuring simulation accuracy. In this paper, a newly method for the model optimization of whole-core fuel-assembly bowing was proposed. The measurement values and theoretical whole-core model are combined in this method based on the 3D variational (3DVAR) and artificial neutral network (ANN) algorithm. In 3DVAR, the cost function is constructed to characterize the difference between the actual fuel-assembly bowing and theoretical model. The 3DVAR algorithm was applied to model optimization and the ANN algorithm was applied to establish the relationship between the whole-core model of fuel-assembly bowing and corresponding simulation of measurement values. For method verifications, the power-distribution measurements of a commercial NPP in China has been applied, and the model that need to be optimized was taken as the bowing amplitude of each fuel-assembly. The numerical results indicate that the relative errors of the power distribution in entire cycle could be reduced notably to meet the industrial criteria, and the maximum value is reduced from 13.4% to 7.7%. As conclude, the proposed method based on 3DVAR algorithm and the measurement values is efficient to optimize the model of fuel-assembly bowing.
Effect of Fission Products on Equilibrium U-Pu and Th-U Cycles for 16 Selected Reactors
Authors: Sabol Josef , Borys Francis , Křepel Jiří 
Published: November 2024

Abstract
Three papers related to the equilibrium fuel cycle in 16 selected reactors were submitted to ICONE 31 conference. In this second part, the impact of fission products on the equilibrium is discussed.
For each of the 16 reactors in the U-Pu and Th-U cycle and in the infinite lattice, the once-through depletion was calculated with a specific share of fission products. During the calculation, the total number of actinides was kept in balance in a closed cycle, only the proportion of fission products grew. In the second step, two possible approximations for multi-batch operation were compared: linear theory of reactivity and equilibrium approximation.
It was found that the equilibrium does not differ so much with burnup, significant changes were seen only in thermal systems in the U-Pu cycle. Since actinides were kept in equilibrium, only a decrease in reactivity was observed, which was purely linear for fast systems and nonlinear for thermal systems. When comparing both approximations, it turned out that they do not fit. In all cases, the equilibrium approximation stabilizes the different concentrations of long-lived and stable fission products, which results in a decrease in parasitic absorption and in an increase in reactivity. For this reason, the paper further analyses the concentration of selected fission products between each other.
All results were obtained using the EQL0D v4-procedure and with ENDF/B-VIII.0 nuclear data library.
Numerical Study of Gas Flow and Diffusion Driven by Buoyancy in a Multi-Openings Compartment
Authors: Jiang Jinghua , Hu Peizheng , Tong Lili , Cao Xuewu 
Published: November 2024

Abstract
During a severe accident in a pressurized water reactor, if the coolant pipe ruptures in a multi-openings compartment, a mixture of hydrogen, vapor, and air migrates and diffuses between compartments, driven by density disparities during the extended period following the cessation of coolant rapid discharge. The same phenomenon occurs during a “wet bypass” in a fusion reactor, a mixture of hydrogen, tritium (HT/HTO), vapor, and air migrates and diffuses between the suppression system, vacuum vessel, and Tokamak gallery. In the scenarios mentioned above, the buoyancy-driven flow resulting from the mixing of light and heavy gases will influence the long-term hydrogen risk and migration of radioactive materials. Previous research has primarily focused on studying the migration behavior of multi-component gas mixtures propelled by jet flow. An experimental device (A4Q-FB) with a multi-openings container was established, to clarify the influence of buoyancy-driven flow on gas migration. The test is carried out with well-controlled boundary conditions, in which, a mixture of helium and vapor with a mass fraction of 2:15 is injected through a vertically downward nozzle into the compartment filled with air. The injection was stopped after 11200 s from the beginning of the experiment, and other boundary conditions were controlled to be unchanged, and the data of the temperature and concentration of the gases inside the container and the flow velocity of the gas mixture in the openings were obtained. In this study, a three-dimensional analysis model for this test was developed, and the predictive accuracy for gas mixture migration and diffusion behavior was assessed. The results indicate a fundamental alignment between the simulation outcomes and the trends observed in the experimental data. This consistency suggests the reliability of the model in the macro-level predictions of phenomena and supports our understanding of the gas mixture migration under buoyancy-driven flow. The simulation results indicate a maximum relative error of 12% in predicting the concentration of vapor at the top and a maximum relative error of 20% in estimating the gas flow velocity within the top openings. The simulation also provides the 3D distribution of velocity and concentration in both container and openings.