bluuebunny/crossref_metadata_2025_split huggingface dataset:
PID Parameter Tuning Method for Steam Generator Level Control Based on Ant Colony Algorithm
Authors: Wang Yulong , Wei Xinyu , Sun PeiWei 
Published: November 2024

Abstract
Steam generator levels in pressurized water reactor nuclear power plants at home and abroad use the classical serial PID control scheme. At present for the steam generator level control system control strategy method research has been fruitful, but when the operating conditions change, the original PID control effect may no longer meet the demand, so the need for PID control parameter calibration of the problem is still very few relevant methods of research. For this reason, this journal proposes a method to train the RBF neural network model based on the historical data of the system, to improve the optimization data, and establish an accurate mathematical model of the system based on the optimized data by the least squares method for the PID parameter tuning method of the subsequent ACO algorithm. The RBF neural network is trained through historical data to obtain the feature model. The optimized data is modelled using F-test to determine the order of the system data, and the corresponding differential equations are used for identification, and the identified differential equations are transformed into transfer functions, through which an accurate mathematical model can be identified in the case that the order of the system structure is not clear. Finally, the ACO algorithm is used to adjust the PID parameters according to the identified mathematical model of the system, and the adjusted PID parameters have a good control effect in different working conditions, and the results show that the optimized data identification transfer function is basically the same as the actual system, and the parameter-optimized control system of the steam generator level control system is better than the original control system in terms of the overshooting control indexes, and it achieves the control of the liquid level inside the steam generator under different working conditions. Steam generator level control under different operating conditions. It is concluded that when the model parameters are changed, the PID parameter control effect of the identification model adjusted by the ACO algorithm is better than the original PID control of the steam generator.
Experimental Study of Heat Transfer Characteristics During Reflooding Process Under Partially Exposed Conditions in Narrow Rectangular Channels
Authors: Wu Junchen , Liu Haidong , Chen Deqi , Zuo Qianlong , Liu Hanzhou , Deng Jian , Wu Dan , Chen Mingjing 
Published: November 2024

Abstract
In the event of a small and medium-sized breach accident in the primary loop of a nuclear reactor, the pressure relief and discharge causes a loss of coolant flow in the core, exposing the top of the fuel elements to overheated steam while the bottom remains submerged in residual cooling water. At this point, the fuel elements experience significant axial temperature differences, making the boiling two-phase flow behavior during the reflooding process of the fuel elements more complex compared to complete exposure. This experiment conducted reflooding tests with varying initial water levels for localized exposure, and the results showed that the minimum wall film boiling temperature in the completely exposed state gradually decreased downstream along the axial direction, while it increased in the case of localized exposure. The precursor cooling duration experienced during the reflooding process was longer for complete exposure than for localized exposure, and at the same wall superheat, the heat transfer coefficient for localized exposure was larger. The average heat transfer coefficient during the reflooding process of localized exposure was higher than for complete exposure, and as the degree of exposure decreased, the average heat transfer coefficient during the precursor cooling phase increased.
Total Monte Carlo Uncertainty Analysis of VVER-440 Spent Nuclear Fuel in LWR and HWR Reactor Environment
Authors: Lovecký Martin , Závorka Jiří , Kořínek Tomáš , Škoda Radek 
Published: November 2024

Abstract
Total Monte Carlo (TMC) extends the Monte Carlo method, using stochastic techniques and random sampling to solve the Boltzmann transport equation in spent nuclear fuel (SNF) depletion analysis. TMC evaluates the impact of uncertainties in nuclear data, such as cross-sections and fission product yields, on SNF characteristics, focusing on decay heat, which is crucial for SNF handling and management. TMC generates random nuclear data variations within uncertainty ranges for Monte Carlo simulations, resulting in outcome distributions (e.g., decay heat) that reflect real-world behavior uncertainties.
The uncertainty analysis examined two reactor environments: a standard VVER-440 reactor using a light water coolant at 1375 MW thermal power and a Teplator district heating reactor VVER-440 fuel in a HWR environment at 50 MW thermal power. This comparison offers insights into VVER-440 SNF behavior under different reactor conditions and coolants.
Using the Serpent 2 code and referencing the ENDF/B-VIII.0 and TENDL nuclear data libraries, the study focused on the impact of variable cross-sections on critical nuclides and fission product yield variations. It analyzed decay heat generation immediately post-reactor shutdown and long-term decay heat for spent fuel cask loading, providing a comprehensive view of the nuclear fuel cycle.
Preliminary Assessment of CATHARE-MODELICA Codes Coupling for Safety Analyses in Nuclear Cogeneration
Authors: De Angelis Alessandro , Olita Paolo , Ambrosini Walter 
Published: November 2024

Abstract
To assure a viable energy transition towards a low-carbon future energy mix, Small Modular Reactors have gained attention in the last years due to their peculiarities, e.g., potential enhanced safety, economic viability due to series production, ease of installation and modularity. Moreover, their implementation in Hybrid Energy Systems in cogeneration operation could bring to the decarbonization of other carbon-emitter industrial sectors. However, the introduction of Small Modular Reactors in Hybrid Energy Systems for cogeneration can result in different transients induced by the thermal users on the nuclear power plant, which must be carefully considered when performing a safety assessment. To this purpose, the paper describes the methodology adopted to perform a code coupling useful for future safety analyses of these hybrid energy systems. In particular, the coupling is performed between a well-assessed thermal hydraulic system code used to perform nuclear safety analyses, namely CATHARE 3, and a widely-adopted a-causal dynamic modelling language, namely MODELICA, exploiting the features of both modelling codes. The results, although preliminary, show good coupling capabilities between the two codes, providing an excellent starting point to perform future analyses of some transients induced on nuclear systems in cogeneration mode.
Study on the Development Path of AFR Storage for Spent Fuel in China
Authors: Shi Lei , Lu Haoran , Wang Yihan , Hu Jian 
Published: November 2024

Abstract
Away-from-reactor (AFR) storage of spent fuel is the “intermediate storage” after transported from a nuclear power plant site before final treatment and disposal. AFR storage not only ensures the safe operation of nuclear power plants, but also provides a buffer for spent fuel reprocessing or geological disposal. It is considered an important industry sector in spent fuel management in various nuclear power countries around the world. Under the background of the “carbon peak and carbon neutrality” goal, China nuclear energy will maintain a steady and rapid pace of development. The annual production of spent fuel will also increase year by year, and the demand for AFR storage will become increasingly large. However, there are various types of pressurized water reactor nuclear power units in China, resulting in different requirements for spent fuel management. Therefore, it is necessary to strengthen top-level design, taking into account many factors such as safety, technological maturity, and economy, and combine with the demand forecast of China’s nuclear power development to develop an AFR storage capacity construction plan. This study shows that it is necessary to adhere to the principle of “more concentration and less dispersion”, and start the construction of China’s spent fuel AFR storage capacity as soon as possible.
Experimental Investigation of Fluidelastic Instability in In-Line Tube Arrays
Authors: Wang Yuqi , Wang Xiaoxin , Shi Li , Wu Xinxin 
Published: November 2024

Abstract
The Helical Tube intermediate heat exchanger (Helical Tube IHX) is a crucial component for process heat applications and hydrogen production in Very High-Temperature Gas-Cooled Reactors (VHTRs). Due to the coil design with a large spiral diameter in the Helical Tube IHX, the fluidelastic instability needs to be given special attention. The simplified straight tube bundle model and inclined tube bundle model were designed and constructed to inclined study the fluidelastic instability of single flexible tube in the center of the bundle. A variable magnetic eddy current damping system has been constructed and utilized to change the tube’s damping ratio. The critical flow velocities for fluidelastic instability in tube bundle model were obtained under different mass damping parameters. By comparing the instability maps of the straight tube bundle and the inclined tube bundle, the results indicate that the inclined tube bundle is more prone to instability compared to the straight tube bundle.
Technology Review and Safety Assessment of Nuclear-Renewable Hybrid Energy Systems With Light-Water Small Modular Reactors
Authors: De Angelis Alessandro , Frignani Michele , Pucciarelli Andrea , Sevbo Oleksandr , Lapuente Miriam Minchole , Schneidesch Christophe , Vaglio-Gaudard Claire , Miss Joachim , Hollands Thorsten , Ambrosini Walter 
Published: November 2024

Abstract
An increase of renewable energy share is expected in the next future to drive the energy transition towards a low-carbon future energy mix. However, renewable energy sources are intermittent and non-dispatchable by nature and could lead to grid stability issues. For this reason, the idea of coupling renewable energy sources with Nuclear Energy (in particular, Small Modular Reactors) in the so-called Nuclear-Renewable Hybrid Energy Systems has gained more attention during the past years. During the second half of 2022 a European project belonging to the Horizon Europe program, namely TANDEM, has started to investigate the safety aspects of Light-Water Small Modular Reactors when introduced in a Hybrid System scenario. This paper aims at summarizing some of the first outcomes of the TANDEM project regarding the technology assessment of a Nuclear-Renewable Hybrid Energy System with Light-Water Small Modular Reactors, with preliminary considerations about safety of nuclear cogeneration. To this purpose, the main components of a general Hybrid system are considered, i.e., Light-Water Small Modular Reactor, low and high temperature hydrogen production, thermal and electrical storages and seawater desalination, aiming at providing general rules to perform the future modelling of the system in the frame of the project.
Experiment and Simulation Analysis for Performance of Squat Shear Wall in Nuclear Island
Authors: Wang Shaojie , Wang Dongmei , Li Yumin 
Published: November 2024

Abstract
In nuclear island, shear wall has the characteristic of low shear span ratio (low aspect ratio) and high reinforcement ratio, which can be called as the squat shear wall. The performance of squat shear wall determines the seismic PSA analysis conclusion of nuclear island structures. According to the typical wall in nuclear island, three large-scale specimens of squat shear wall are designed. Considering different axial compressions, quasi-static experiments are conducted. The load, displacement, cracks and final failure mode are obtained. Hysteresis curve, skeleton curve, energy dissipation capacity, ductility coefficient, stiffness degradation and other performance data are calculated and analyzed. A proper nonlinear simulation method is presented where the key parameters in simulation are calibrated by experimental data. The FEM results and experiment results are in good agreements. The result shows that the squat shear wall in nuclear island has high loading bearing capacity, but the ability to resist deformation is limited, so the wall tends to be brittle. The reduction of axial compression ratio can enhance the energy dissipation capacity of shear walls. The variation of axial compression ratio has little impact on the stiffness degradation.
Study on Stabilization Technology of Base Isolation Layer Using Air Floating Technology by Shaking Table Test
Authors: Takuto Aramaki , Furuya Osamu , Hiyama Yoshiro , Yamazaki Koji 
Published: November 2024

Abstract
Japan is one of the world’s most earthquake-prone countries, and in facilities dealing with ultra-fine processing equipment, there is a focus on maintaining system functionality during earthquakes below a certain magnitude to prevent facility damage during major earthquakes. This emphasis aims to reduce downtime and production losses by minimizing equipment stoppages and facilitating restarts after earthquakes below a certain magnitude. To meet these performance requirements, the development of new seismic isolation technologies is being sought. While air floating units exhibit high vibration isolation performance against horizontal input, concerns have been raised about the stability of load support performance in the vertical direction [1,6]. To address this, additional spring elements have been introduced to maintain the load support performance of the air floating unit. However, there are concerns that the introduction of spring action may worsen the vibration conditions of the seismic isolation target due to rocking vibrations. This paper introduces additional spring elements to suppress rocking vibrations on the upper part of the air levitation unit and presents the performance evaluation results from vibration table experiments. As a result of adjusting the additional spring, it was confirmed that rocking vibrations could be suppressed even during simultaneous vertical and horizontal input, maintaining horizontal seismic performance. Additionally, the analytical model was validated against the experimental results, confirming the validity of the analytical model.
Study on the Characteristics of Nuclear Fuel Industry
Authors: Shi Lei , Wang Yihan , Liu Hongjun , Zhang Honglin 
Published: November 2024

Abstract
The nuclear fuel industry is related to high-tech strategy and energy security. It is an important technical and material foundation for China’s nuclear power development. In the new era, in order to better promote the high-quality development of China’s nuclear fuel industry, it is necessary to summarize historical experience, further clarify the path of development, carefully study and deeply grasp the characteristics of the industry itself, and on this basis, provide correct guidance for the development of China’s nuclear fuel industry. This study summarizes the development experience of the nuclear fuel industry in major countries, and objectively analyzes the characteristics of the nuclear fuel industry from multiple aspects such as country, technology, enterprise, and market. Study has shown that nuclear fuel has the characteristics of a “pyramid” structure, with the top-level features of nuclear deterrence and international nuclear safeguards.
Application of Virtual Reality Technology in Human Factors Verification of Nuclear Power Plant Design
Authors: Hao Dong , Fan YanFang , Bai BingHe 
Published: November 2024

Abstract
This paper introduces a Virtual Reality (VR)-based Human Factors Engineering (HFE) framework for enhancing nuclear power plant designs through quantitative assessments, employing VR simulations of construction and maintenance to reduce design revisions. The initial phase entailed the creation of an integrative tool leveraging Unity and C#, which substantially mitigates the complexities associated with migrating from conventional modeling environments to VR visualizations. Subsequent stages involved the adoption of advanced virtual design methodologies and motion capture technologies to systematically capture and analyze human kinematics and interactions within simulated settings. Following this, a multi-dimensional evaluation system has been customized specifically for the nuclear environment, including operability, comfort, special environments. The fuzzy comprehensive evaluation method is used to conduct quantitative evaluation based on the established evaluation system and the data captured by the system. This methodology enables the anticipation of potential ergonomic issues at the design stage, allowing for iterative improvements and the generation of optimized solutions. Furthermore, the paper exemplifies the practical application of this method through two case studies: the selection of steel cage joints during construction and the optimization of evaporator compartment layouts during maintenance, illustrating the method’s efficacy. This comprehensive approach is designed to be applicable across the entire spectrum of design phases, from conceptualization to detailed execution, thereby addressing the distinctive challenges inherent in nuclear power plant engineering. Collectively, this research contributes to the enhancement of design quality, cost-efficiency, and operational efficacy in the field of nuclear power plant design.
Development and Optimization Criteria of Cementitious Mortars Used for the Immobilization of ILW Radioactive Waste
Authors: Rizzo Francesco , Rosa Domenico , Mangialardi Teresa , Di Palma Luca , Giannetti Fabio 
Published: November 2024

Abstract
The cementing process is the most widely adopted method of intermediate-level waste (ILW) conditioning, involving the use of cement mortars to waste incorporation. The main challenge in processing these mortars lies in achieving an optimal balance among their components. Accordingly, this study aims to investigate the essential criteria for developing prototype cement mortars suitable for immobilizing Intermediate Level Waste by investigating the effect of key parameters, such as water/cement and sand/cement ratios, on the workability of the compound. Different types of cement (CEM I, CEM II and CEM IV) were used, also analyzing the effect of aggregate particle size composition on workability, and the benefits of adding superplasticizer to reduce the water/cement ratio and ensure high mechanical strength were evaluated.
It was observed that increasing the water/cement ratio enhanced workability, while increasing the sand/cement ratio decreased it. The use of superplasticizer decreased the water/cement ratio with the same workability. In addition, the grain size composition like that provided by Andreasen improved workability compared with the as-is composition. The investigation and the integrated multifactorial approach allow for a rapid convergence towards the production of a prototype mortar with the desired characteristics for the immobilization of Intermediate Level Waste.
A Stabilization Strategy of Multi-Resolution Multiphase MPS Method
Authors: Zhong Yubao , Li Sijun , Chen Ronghua , Guo Kailun , Tian Wenxi 
Published: November 2024

Abstract
CFD simulations of multiphase behavior are essential in the safety analysis of nuclear reactors. As a particle method, the MPS method inherently possesses the capability of capturing interfaces or free surfaces. The PArticle method for Nuclear reactor thermal hyDraulic Analysis (PANDA) is a software platform based on particle methods, aiming to provide a safety analysis of key thermal-hydraulic phenomenon in nuclear reactors. The integrated PANDA-BD adopts the specially designed Multiphase MPS method (MMPS method), and shows great ability in difficult multiphase simulation.
However, due to the semi-implicit algorithm and the uniformity of particle size, MMPS incurs a relatively high computational cost. To improve the efficiency, some multi-resolution techniques have been developed for the MPS method.
Another challenge with the MPS method could be the stabilizing issue. A seriously bias particle distribution could lead to the failure of the simulation. Most of the current rearrangement methods assume a uniform particle size, and are therefore not suitable under a multi-resolution scheme.
In this study, the concept of local compressed ratio (LCR) is proposed to measure the uniformity of the particle distribution under multi-resolution. An optimized particle-shifting technique based on the novel interface recognition and LCR is developed. A series of regularization tests are performed, showing the regularization ability under both single and multiple resolution cases. Simulations of Rayleigh-Taylor Instability are presented, the novel interface detection is tested and shows good accuracy, the robustness and efficiency of the proposed method for multiphase simulation is proven.
Application of Global Importance Measures in RISMC for Dynamic Risk Assessment
Authors: Ma Liye , Wang He , Wang Longcong 
Published: November 2024

Abstract
Risk-Informed Safety Margin Characterization (RISMC) integrates traditional deterministic safety analysis methods with probabilistic safety analysis approaches by employing stochastic sampling of uncertain parameters and dynamic branching simulations of nuclear power plants. It enables time-dependent, uncertainty-aware, and multi-level risk assessments. In the process of RISMC analysis, quantifying the impacts of uncertainty parameters, response events, and system components on nuclear power plant risk can support the implementation of risk analysis results in nuclear power plants. However, traditional importance measurement methods cannot directly apply to RISMC. This study adopts the global importance measurement method to measure the time-dependent importance. The selected global importance measurement method considers input parameter uncertainties and uncertainties from interactions between input parameters. It is demonstrated on Station Blackout (SBO) accidents modeled with dynamic event trees and compared with traditional importance metrics used in probabilistic risk assessment. The risk importance of systems and physical design parameters in SBO accidents is ranked. Results indicate timely AC power recovery significantly reduces risk, while the importance of TAFW failure diminishes over time. Future research is needed to investigate the discrepancies between the global importance measurement method and traditional methods.
Modelling Mass Transfer and Mixing of Corium with Concrete during Molten Corium-Concrete Interaction: Impact of Concrete Thermal Conductivity
Authors: Khurshid Ilyas , Addad Yacine , Afgan Imran 
Published: November 2024

Abstract
Molten corium, a mixture of nuclear fuel, claddings, and structural elements, would form after the meltdown of a reactor core in the event of a severe accident at a nuclear power plant. This paper presents the development and implementation of a new numerical mixing model in OpenFOAM for a comprehensive analysis of concrete ablation during the Molten Corium-Concrete Interaction (MCCI) process. The well-researched CCI-2 experiment served as the model’s benchmark during model validation. The model can forecast heat transfer, concrete ablation, phase change, mass transfer, and mixing of concrete and corium. Additionally, the effect of concrete thermal conductivity was investigated to study its impact on concrete ablation. We observed significant low concrete ablation, controlled temperature, velocity and density field for concrete with low thermal conductivity. Accordingly, the ablation of concrete decreased by 61 and 58 %, when the thermal conductivity was decreased to 1.5 and 1.0 W/m.K, respectively from 2.0 W/m.K. Our results also revealed that the melted concrete and corium, even though, having different densities, form a single phase/mixer that could enhances the ablation of concrete.