bluuebunny/crossref_metadata_2025_split huggingface dataset:
Validation and Verification of ASYST Code for Predicting Condensation Phenomena in Nuclear Reactor Safety Systems
Authors: Saraswat Satya Prakash , Hassan Mubashir , Osman Sameer Mohammad , Allison Chris , Addad Yacine 
Published: November 2024

Abstract
The ASYST (Adaptive System Thermal-Hydraulic) code series is a recent development that integrates the functionalities of SCDAPSIM and SAMPSON. It can simulate and analyze fluid and reactor systems with Best Estimate Plus Uncertainty (BEPU) capability. The proposed work involves conducting a thorough validation and verification process for the ASYST code model to predict the condensation phenomena using a condensation-separate effect test facility (present at the Department of Mechanical and Nuclear Engineering, Khalifa University), specifically focusing on the condensation phase. The condensation separate effect test’s experimental facility was devised by members of the thermal-hydraulic team affiliated with the Nuclear Engineering Department at Khalifa University’s Emirates Nuclear Technology Centre. This apparatus is specifically configured to acquire dependable data on condensation heat transfer analysis in a vertical tube simulating the condensing element within a Passive Containment Cooling System (PCCS), wherein steam and non-condensable gas mixtures are examined. The aim is to provide a detailed description of how the ASYST code aligns with and accurately represents the outcomes of the test facility.
The condensation of the separate effect test facility is a critical benchmark for this validation and verification framework. By scrutinizing the code’s performance concerning the test facility outcomes, this approach comprehensively evaluates the ASYST code’s reliability and applicability to practical scenarios of nuclear reactors’ passive safety systems.
RCCA Ejection Accident Analysis Based on Coupled 3d Neutronics and Thermal-Hydraulics
Authors: Liye Zang 
Published: November 2024

Abstract
The rod cluster control assembly (RCCA) ejection accident is one of the most important Design Basis Condition (DBC) of Pressurized Water Reactor (PWR) nuclear power plant. This accident leads to a rapid reactivity transient by inducing an uncontrolled positive reactivity, followed by a nuclear power increase and a power distribution disturbance. The nuclear power increase is limited by reactivity feedback effects, such as the Doppler feedback effect caused by the fuel temperature increase. This may result in Departure from Nucleate Boiling (DNB) among the adjacent fuel channels, and the potential failure for the fuel cladding.
At present, the traditional decoupling analysis method is still used in RCCA ejection accident (REA) analysis of HPR1000, which has a high degree of design conservatism and restricts the continuous improvement of the unit’s economy and performance improvement. This paper takes a certain HPR1000 reactor as the research object, establishes a coupled 3D neutronics and thermal-hydraulics analysis method based on SMART/FLICA III-F code. The analysis process and results were compared with traditional method. The research results indicate that the coupled 3D neutronics and thermal-hydraulics analysis method is reasonable and feasible, which can more realistically simulate the core situation during the REA transient process, and can effectively release the conservatism of key parameters. On the basis of ensuring the same level of safety, it provides the possibility for improving the performance of the unit.
Simulation of Aerosol Transport Under Spray Control Measures for Floating Reactor Accidents
Authors: Liu Zhe , Wang Ning , Yang Ya-Peng 
Published: November 2024

Abstract
A three-dimensional accurate simulation was conducted using computational fluid dynamics methods for the spray removal of aerosols after a large break accident in a floating reactor. The Euler Euler Lagrange method was used to simulate gas solid liquid, and grid independence work was carried out to verify the reliability of the grid model. The physical model adopted takes into account the inertial collision, interception, and diffusion effects of droplets on aerosol particles. The simulation results can accurately provide the flow field and droplet motion path distribution in the cabin. At the same time, the removal effects of different mechanisms on aerosols, as well as the removal effects of single droplets and droplet groups on aerosols, are compared.
Investigation on Isotopic Depletion Validation and Uncertainty Analysis With Kalman Filtering
Authors: Qiao Jianshu , Huang Dongli 
Published: November 2024

Abstract
Isotopic concentration is one of the essential indicators to evaluate the burnup process, which provides detailed descriptions of fuel assembly status among various reactor core conditions. This work aims to develop a method in support of isotopic depletion validation. The developed validation method attempts to improve predictions of isotopic concentration across burnup fusing the computational results from a high-fidelity model with a small amount of measurement data. The isotopic predictions are improved by minimizing the discrepancies between the model prediction and the reality, as well as reducing the associated predicting uncertainties. However, this is such a challenge due to the lack of experimental measurement data and limitations by qualities of existing data. Isotopic validation method developed in this work takes advantages of Kalman Filtering (KF) to solve the challenges mentioned above. This work makes the best use of the reduced data to optimize the prediction process based on a few measurements across burnup. The simulations and uncertainty analysis are completed by OpenMC with a buffer code. Results show that the discrepancies between computational predictions and the reality are reduced, and uncertainties of isotopic concentrations across burnup decrease by about 50%.
Analysis of Energy Dissipation Effect of Fluid-Filled Capacitive Structure in the Water Hammer Process
Authors: Yang Linqing , Qin Benke , Li YanLin , Ma Yue , Bo Hanliang 
Published: November 2024

Abstract
Water hammer is a common phenomenon in nuclear engineering piping systems. Water hammer can cause a large fluctuation of fluid pressure, which may threaten system security. It is essential to control and mitigate pressure fluctuations and energy. The fluid-filled capacitive structure is a key component of energy dissipation equipment, and plays an important role in the water hammer mitigation process. The water hammer performance experiment rig was established, and the experiments were carried out. The water hammer theoretical model was established, and the influence of unsteady friction was considered. The model results agreed well with the experimental data. The fluid-filled capacitive structure was added to the pipeline system to analyze the dissipation effects using two analysis methods. The analysis results show that as the fluid-filled capacitive structure moves from upstream to downstream and the cavity volume increases, the fluctuation amplitude of pressure and internal energy decrease, the fluid mass transfer between the capacitive structure and the pipes is enhanced, and the energy dissipation effect increases. The capacitive structure periodically absorbs and releases the energy of the water hammer process to achieve the energy dissipation effect. The research results can provide guidance for the design of water hammer pressure mitigation and energy dissipation equipment in nuclear engineering pipeline systems.
CFD Analysis of Heat Transfer in Molten Salt Fuel Chambers of the Wielenga Innovation Static Salt Reactor (WISSR)
Authors: Khaleb Ikhwan , Wielenga Thomas , Yang Won Sik 
Published: November 2024

Abstract
This paper presents the computational fluid dynamics (CFD) analysis to study the heat transfer in the Wielenga Innovation Static Salt Reactor (WISSR), which is a fast-spectrum TRU-burner reactor. The core consists of thin annular fuel chambers holding NaCl-(TRU,U)Cl3 fuel salts that are arranged concentrically, and NaCl-KCl-MgCl2 coolant flowing through the gaps between fuel chambers removes the heat. In the central fuel chambers, stainless steel corrugations are used to enhance the conduction of heat from fuel to coolant. The CFD analysis focuses on two main aspects. Firstly, it assesses the natural circulation of fuel salt. Secondly, the interaction of fuel salt and helium gas at the top of chambers is investigated using the volume of fluid (VOF) method. The CFD analysis confirms the feasibility of the design.
Numerical Investigation of Postulated Steam Generator Tube Rupture Accident in a Lead-Cooled Fast Reactor
Authors: Chen Yutong , Zhang Dalin , Feng Zhenyu , Lin Yue , Tian Wenxi , Qiu Suizheng , Su Guanghui 
Published: November 2024

Abstract
Because of the unique design of Lead-cooled Fast Reactor (LFR), the steam generator tubes containing high-pressure water is placed directly in the high temperature lead pool, therefore once a steam generator water tube ruptures during operation, the high-pressure subcooled feed water would interact directly with hot lead, giving rise to Steam Generator Tube Rupture (SGTR) accident, which is one of the accident scenarios that must be considered during LFR design and safety analysis process. Focused on this topic, numerous experimental and theoretical research have been conducted in recent decades. In this study, the Eulerian multiphase analysis code ACENA was applied to simulate the postulated SGTR accident of European pool-type lead-cooled fast reactor ALFRED. Attention was paid to the simulation results of safety parameters, such as the evolution history of maximum cladding and fuel temperature, core relative power, core void fraction and cover gas pressure; the migration process of steam bubbles was visualized as well. Furthermore, the sensitivity of model ruptured tube quantities and neutronic feedback models were also analyzed. This study could provide reference for LFR design and safety analysis work.
Discussion on Radiation Protection for PWR Second-Loop Steam Supply Outside for Industrial Application
Authors: Yaping Zhuang , Yuan Fang , Wu Fang , Miao Zhengqiang 
Published: November 2024

Abstract
Nuclear power units have the ability to provide large-scale, stable industrial steam. With the rising prices of fossil fuels, nuclear energy has become economically competitive in supplying industrial steam. Laboratory measurements show that, under the condition of no leakage in the heat transfer tube of the steam generator, the concentration of radionuclides in the condensed water of the secondary loop steam is not different from that in the environmental water sample except for tritium. The IAEA technical report provides a two-cycle scheme of PWR supply steam for SAGD, which has the advantages of simple system, less process equipment, and high steam quality compared to the external steam supply scheme from the tertiary loop. Based on the drinking water standard, this article proposes to control the target value of radionuclide management for industrial application of secondary loop steam at 3.0 Bq/kg; Add a high-sensitivity radiation monitor on the steam generator blowdown line, and carry out technical renovation on the main steam pipeline N-16 radiation monitoring channel to provide timely, effective, and reliable monitoring of steam generator heat transfer tube leakage. It is interlocked with the control valve at the boundary of the external steam supply pipeline factory area to prevent radionuclides from passing through the steam supply pipeline factory area during operation and accident conditions.
Development of a New Crust Model for Analyzing VULCANO VBS-U3 MCCI Experiment With MPS Method
Authors: Yamada Takeshi , Li Xin , Yamashita Takuya , Yamaji Akifumi 
Published: November 2024

Abstract
In the postulated severe accident of a light water reactor, the molten core material (corium) may be released to the reactor containment floor and the Molten Core - Concrete Interaction (MCCI) may take place. This complex phenomena may involve dehydration, thermal decomposition, and ablation of concrete, redox reactions of oxidic and metallic phases of corium, solidification and remelting of corium, mixing of molten concrete with corium and so on. Evaluation of the resulting concrete floor ablation is important as it may threaten the containment boundary.
The Moving Particle Semi-implicit (MPS) method discretizes fluids and solids with calculation points (particles) in a Lagrangian manner and has been extensively developed for analyses of various MCCI tests with consideration of crust formation at the melt-concrete interface. Due to its Lagrangian nature, there is always some randomness to each particle position and Particle Shifting (PS) is one of the numerical techniques, which corrects such randomness by slightly shifting the particle position from a particle dense region to a coarse region to sustain numerical accuracy of the MPS method. Such slight correction does not usually have any physical impacts on the calculation unless the particle velocity becomes extremely small and duration of the phenomena becomes extremely long.
In this study, a new crust model is being developed to analyze MCCI, which involves continuous concrete ablation with presence of the crust layer between the corium and the concrete walls, which may gradually move with the slow concrete wall ablation process over long time. The new crust model must enable accumulation of physical displacement of the crust particle over long time (i.e., enable physical creeping) while preventing accumulation of numerical displacement of the crust particles over long time (i.e., preventing numerical creeping), Hence, in the new crust model, the PS has been effectively disabled for the crust particles. Qualitative validity of such numerical modeling was confirmed through some trial analyses of VULCANO-VBS test using a set of tentative calculation conditions and parameters, which should be carefully revised for future quantitative discussions including validation of the analysis results with experimental results.
Control Study for Automatic Startup of a Pressurized Water Reactor Nuclear Power Plant
Authors: Zhang Qi , Yu Wenhao , Xiao Longhao , Sun PeiWei , Wei Xinyu 
Published: November 2024

Abstract
To improve the automation for the startup process of a pressurizer water reactor (PWR) nuclear power plant (NPP), reduce the work intensity of the reactor operator, shorten the startup time, and improve the correctness and standardization of the process, the automatic control of NPP startup process is studied in this paper. First, the control scope and the requirements of automatic startup control system are analyzed and studied based on the characteristics and the operation management process of PWR NPP. Aiming at the current manual control tasks, three realization control methods are proposed, including sequence control, traditional closed loop control and advanced algorithm. A simulation platform for automatic startup system of typical PWR NPP was established based on 3KeyMaster, Relap5, and MATLAB/Simulink. Two typical manual control processes in state 3 for the startup process are selected to design the automatic control system based on PI controller and fuzzy controller respectively, and the simulation tests are carried out. The simulation results show that the designed control systems can realize the automatic control of the processes. Compared with the manual control, the control performance is improved, and the operation steps and work burden of operator are reduced. This study can provide a reference for the design of automatic startup and shutdown control system, and it is of great significance for improving the automation of startup and shutdown process of PWR NPP.
Stress Intensity Factor Solutions for Circumferential Through-Wall Cracks Applicable to Pool Type Sodium Cooled Fast Reactors
Authors: Yada Hiroki , Takaya Shigeru , Machida Hideo 
Published: November 2024

Abstract
ASME Code Case N-875 provides a rational in-service inspection (ISI) approach for liquid-metal cooled reactors. This is a risk informed approach and ISI requirements including acceptance criteria are established by considering features of respective plants in terms of the effects of flaws on the plant safety. In this approach, fracture mechanics is essential. Not only part-through-wall cracks but also through-wall cracks need to be evaluated, for example, to determine the maximum allowable size of flaws in reactor internal components and investigate the applicability of continuous leakage monitoring to flaws in sodium-retaining components. The basic procedure of the Code Case has also been incorporated in Fitness-for-service code (Section XI, Div. 2) in ASME, which provides requirements for Reliability and Integrity Management programs for nuclear power plants, including advanced reactors. The demonstration sodium-cooled fast reactor currently under development in Japan is expected to be designed with thin wall and large diameter to reduce the thermal stress. For some components, the ratio of radius to thickness is expected to exceed 100. There are currently no generalized Stress Intensity Factors (SIFs), which are required for fracture mechanics, applicable to components with such a large ratio of radius to thickness. In this study, as a part of the development of a flaw evaluation method applicable to components with large ratio of radius to thickness, the conservatism of applying the SIF solutions for a through-wall crack in a plate to a circumferential through-wall crack in a cylinder with large ratio of radius to thickness was discussed. As a result, it was clarified that the SIF solution for a plate should not be used for circumferential through-wall cracks. Therefore, a new SIF solution of a circumferential through-wall crack in a cylinder was developed by using Finite Element Analysis by Abaqus.
Study on Neutronics Properties of Zirconium Hydride Moderator in Reactor
Authors: Shixin Lin , Zheng Youqi 
Published: November 2024

Abstract
The metal hydride ZrH1.6 has a high hydrogen atom density, equivalent to that of liquid water. Compared to other hydrides (such as H2O and organic compounds), particularly at high temperatures, ZrH1.6 exhibits good strength and stability. Using hydrogen zirconium in a reactor can effectively slow down fast neutrons, thereby increasing the utilization of thermal neutrons and enabling the use of low-enriched uranium fuel. However, as a Solid state moderator matrix, hydrogen zirconium has a relatively large positive temperature coefficient in a reactor, posing a threat to the inherent safety of the reactor.
The goal of this study was to find out why the temperature coefficients of zirconium hydride is positive and to determine when the moderator and fuel temperature coefficients become positive. These are analyzed in this article by simplifing fuel and moderator calculation models and applying the SARAX program, from the most fundamental physical theory, the neutron physics parameters such as reactivity changes, nuclear cross-sections, neutron energy spectra, and reaction rates of the simplified model before and after temperature rise. The results indicate that the positive temperature coefficient of metal hydride ZrH1.6 is due to the hardening of the neutron energy spectrum below 1eV when the fuel and ZrH1.6 are heated, causing some neutrons to escape from the moderator and enter the fuel, thereby increasing the fission of the fuel. Moreover, when the reactor is designed in the under moderated zone, the temperature coefficient of fuel and moderator can be guaranteed to be negative, thus ensuring the safety of the reactor.
Scaled Experiment on Flow Characteristics of a Single Area of Spent Fuel Dry-Storage System of HTR-PM1000
Authors: Wang Jinhua , Wang Mengyao , Gong Menghang , Li Yue , Wu Bin , Zhang Wei , Gao Lihua , Ma Tao , Wang Haitao , Liu Bing 
Published: November 2024

Abstract
China is promoting the project of a 1000MW high-temperature gas-cooled reactor based on the technology of the pebble-bed module (HTR-PM1000). The spent fuel dry-storage system of HTR-PM1000 is divided into six regions, with each region containing multiple storage silos. The spent fuel canisters are vertically placed in the storage silos, where air is forced or naturally circulated to remove decay heat. Considering the cost of experiments and other factors, a 1/5 scaled experiment was conducted in this paper. The flow distribution and total resistance characteristics of different silos were recorded under the conditions of different fans and canisters loading. In order to eliminate the influence of fan performance on the experiment, two fans were operated at full-load separately, as well as the condition of two fans operating in parallel at full-load. To find the worst case and study the influence of the number of canisters on flow distribution, experiments were conducted for three operating conditions: no-load, full-load, and half-load. The results showed that the trend of silos flow variation was similar between no-load and full-load conditions. Under the half-load condition, the flow decreased with the increase of the number of storage canisters. Additionally, based on the total resistance characteristics, the resistance coefficients under the no-load and full-load conditions were obtained, which matched well with the previously calculated values.
Research on Human Reliability Analysis in Digital Main Control Room of Nuclear Power Plant
Authors: Wang Shuting , Zhao Ling , Liu Dalin , Yan Zhen 
Published: November 2024

Abstract
With the development of computer technology, nuclear power plant instrumentation and control systems have evolved from analog to digital control, and the introduction of new technologies has changed the contextual environment of information display, procedure, and control, thus requiring consideration of new human factors issues that may be brought about as a result. Human Reliability analysis in digital control systems is a key issue related to the safety of nuclear power plants, and there is no precedent for reference or experience to be gained. There is no precedent for reference nor experience to learn from, therefore, how to systematically carry out the human reliability analysis in digital control system is a hot and difficult issue in the current academic and engineering research. This paper discusses the design difference between digital main control room and analog technology main control room with the background of digital nuclear power plant, and at the same time explains how the human reliability changes in the digital environment, classifies the human error modes in the digital main control room system, and carries out a simple qualitative and quantitative analysis of the HRA of digital nuclear power plant based on a specific power plant.
Advanced Radioactive Material Removal System by Silver Zeolite (7) Evaluation of Noble Gas Adsorption Characteristics
Authors: Ishikawa Yoshihiro , Endo Koji , Narabayashi Tadashi , Kawahara Yasuhiro , Nakasaka Yuta 
Published: November 2024

Abstract
It is well known that it is difficult to remove noble gases generated during a nuclear accident. Currently, a noble gas holdup system made from activated carbon is being used. However, activated carbon has many shortcomings, such as flammability, non-moisture resistance, small absorption capacity et al. We have developed a new silver zeolite, “XeA,” as a noble gas adsorbent. The evaluation result shows that the holding time of this new absorbent for xenon is 480 minutes in dry air. This result is 80 times longer than that of activated carbon. Due to its excellent adsorption characteristics, it is expected to replace activated carbon in the vent system.
For this reason, there is a possibility that the volume of the existing system using activated carbon can be reduced to 1/80. Meanwhile, we also propose a system combining this XeA with a filter that does not transmit xenon.