bluuebunny/crossref_metadata_2025_split huggingface dataset:
Reliability Quantification of Passive Containment Cooling System Through Response Surface Methodology
Authors: Shikang Chen , Ronghua Chen , Suizheng Qiu 
Published: November 2024

Abstract
For the new generation of nuclear power plants that widely adopt passive system design, the reliability analysis of passive safety systems shows their important impact increasingly. In addition to hardware failure, functional failure of passive systems is a special and important type of failure modes, which is physical process failure caused by the uncertainty of design functions. This study took the probabilistic safety analysis study of the Passive Containment Cooling System as an example, taking loss of coolant accident as passive system reliability analysis scenario. The functional criteria was based on the pressure within the containment during the long-term cooling period after the PCCS was put into use. The study determined the key operating parameters and design parameters that affect PCCS based on the above scenario and criteria. The study used rank correlation coefficient to accomplish the sensitivity analysis of key parameters on system functions, found out the key influencing factors, and simplified the model. After reducing the number of parameters involved in the study and focusing on key parameters, the artificial neural network (ANN) method was adopted. Based on the best estimation procedure calculation results of design points, the ANN obtained an accurate limit surface with respect to the functional criterion. By quantifying the uncertainty of key parameters, a large number of samples were obtained to complete the response surface calculation, and the failure probability of the system under corresponding accident conditions was obtained. Compared with traditional probability evaluation methods, this research provided a new method to quantify the safety of passive systems, which has higher calculation efficiency and ensured calculation accuracy.
Research on Operation and Control Characteristics of Large Parallel Pipeline Fluid System
Authors: Yuchen Niu , Peng Minjun , Zhang Bowen , Yuan Zhe 
Published: November 2024

Abstract
Large parallel fluid pipeline systems, which exist widely in nuclear power, refrigeration, oil exploitation, and transport as well as other industry, have complex physical coupling relationships among various branches and the main pipeline. This leads to long adjustment cycles of flow and pressure during actual process of conditions changing. Therefore, this study constructed a Modelica language model for an experimental platform of a secondary circuit lubrication system of marine nuclear power plant to investigate rapid adjustment methods for the flow and pressure, thereby enhancing the overall safety and stability. The results indicate that by installing FID controllers on the turbine oil pump and various branch throttling valves, synchronous adjustment of flow and pressure in multiple branches was successfully realized. In comparison to traditional FID controllers, Feedward FID controllers exhibit better resistance to disturbances. FID controllers optimized through the principles of fuzzy control can more rapidly adjust various branch flows to target values, effectively reducing adjustment time. The controller parameter adaptive adjustment scheme can select optimal controller parameters for different condition changing processes.
Design and Performance Verification of Passive Core Cooling System Strainer in Advanced Nuclear Power Plants of China
Authors: Huang Ruolin 
Published: November 2024

Abstract
Once a loss-of-coolant accident (LOCA) occurs in the containment, various materials such as insulation and coating will be damaged by high-energy jet, and the ECCS strainer is used to filter the debris to avoid clogging of downstream equipment. A new advanced nuclear power plant in China equipped with a Passive Core Cooling system (PSI), which relies on the natural circulation driven by the water head difference between sump and core to achieve long-term core cooling after a LOCA. Some pipe break may locate under the sump flood-up elevation, resulting in most of the debris directly entering the core through the break instead of being filtered through the strainer. So a series of debris sources prevention criteria is specified to strictly minimized debris load to avoid clogging the fuel assembly flow path. An ECCS strainer with W-shaped bending filter screen is designed and a strainer module is used as the test prototype to evaluate head loss. According to the test results, 120m2 is selected as the suitable screen area; the chemical precipitate would cause a sharp head loss increase and thus the use of exposed aluminum material should be strictly restricted.
Lattice Physics Study of Various Fuel Assembly Design and Optimization Based on Neutronic Analysis for PT-SCWR
Authors: Shivam Shuvendu , Sekhar Bhogilla Satya , Dutta Goutam 
Published: November 2024

Abstract
The pressure-tube type supercritical water-cooled reactor (PT-SCWR) is a Canadian Generation IV concept advanced reactor designed for higher system pressure and enhanced thermal efficiency. Operating at 25 MPa pressure and temperatures ranging from 350 °C to 625 °C, it achieves over 45% power conversion efficiency. The large variation of coolant density in the core by more than a factor of seven is the driving force for this in-depth core neutronic analysis. Improving lattice parameters in order to achieve the right balance between coolant void reactivity (CVR) and fuel utilization is a major difficulty in the reactor’s physics design. The PT-SCWR features detachable fuel components in pressure tubes immersed in a heavy water moderator. This research focuses on lattice physics and core neutronic analysis, utilizing the DRAGON neutron transport code. Various fuel assembly configurations, including several radial enrichment profiles for different fuels and optimized central rod constituents, were examined for improved safety and efficiency. Innovations, like a two-ring fuel assembly with a central flow tube, enhance fuel exit burnup and reduce coolant void reactivity. The study demonstrates the safety improvements through optimized bundles, meeting criteria for fissile fuel usage and negative coolant reactivity coefficients.
Research on the ODR-VS Method Applicable to Various Geometric Models
Authors: Feng Zhiyuan , Liang Jingang , Guo Wenli , Wang Kan 
Published: November 2024

Abstract
To improve fuel performance and overcome the lack of thermal conductivity and ductility of ceramic nuclear fuel, metal-based fuel elements are used in some high-flux research and test reactors, high-power power reactors and long-life power reactors. However, the stochastic dispersion of fuel particles presents a challenge for accurate modeling and neutron transport analysis. Two commonly employed explicit modeling approaches for stochastic media are the Random Sequential Addition (RSA) method and the Distinct Element Method (DEM). However, the RSA method is constrained by a maximum packing fraction (38.4%). While the DEM method overcomes the limitation of the packing fraction, its inherent complexity, substantial computational requirements, and low packing efficiency remain significant challenges. Moreover, different shapes of fuel elements impose higher demands on geometric universality. Building upon the proposed ODR-VS (Optimized Dropping and Rolling and Virtual Surface) method, we further extend its geometric applicability to include shapes such as cubes, sector cylinders, annular cylinders, and spheres, addressing the geometric modeling needs of a wide range of fuel elements. By employing the enhanced ODR-VS method, we model fuel elements of various shapes and conduct an analysis of the packing efficiency for each type.
Design of a He-Xe Gas-Cooled Nuclear Microreactor System Based on Detailed Models of the Cycle Components
Authors: Guan Chaoran , Chai Xiang , Zhang Tengfei , Liu Xiaojing 
Published: November 2024

Abstract
Nuclear microreactor has disruptive potential as an alternative to carbon-intensive energy technologies based on its mobility, resilience, independence from the grid, long refueling intervals, and low-carbon emissions. A He-Xe gas-cooled nuclear microreactor with the dry-cooled Closed Brayton cycle is designed in this study based on detailed models of the cycle components. Design methods of different components are included. A one-dimensional single-channel model is developed for the reactor core. A one-dimensional discretization is also used for the recuperator, while the two-dimensional temperature field is adopted for the design of the precooler. The turbomachinery uses the mean-line design method. Three He-Xe mixtures are investigated, namely 4 g/mol (pure helium), 15 g/mol, and 40 g/mol. It is found that the system volume is mainly occupied by the reactor core and precooler, and the volume of the turbo-machines is rather small. The 40 g/mol He-Xe achieves the highest thermal efficiency of 42.09% due to the efficiency improvement of the turbo-machines, and the turbomachinery reduces from 5 stages of pure helium to a single-stage of 40 g/mol He-Xe.
Research on Natural Frequency Characteristics of Spiral Tube Bundles
Authors: Li Xiaoxi , Wen Jiming , Gao Puzhen , Yin Yibo 
Published: November 2024

Abstract
Liquid metal reactors (LMR), with the significant advantages in high safety and good economic benefits, has obvious advantages and broad application prospects in the fourth generation nuclear power systems. Spiral tube steam generator is one of the most important equipment in LMR, which are composed of spiral heat transfer tubes, inner and outer cylinders, feedwater headers, steam headers, and other structures. Due to their compact structure and high heat transfer efficiency, they are conducive to equipment miniaturization and have been widely used in different reactors in various countries. Due to the impact of the shell side fluid in the steam generator on the tube bundle, it is inevitable to cause flow induced vibration (FIV), resulting in thinning or even rupture of the tube wall, thereby shortening the service life or even failure of the steam generator. Therefore, the study of the vibration characteristics of spiral tube bundles is of great significance. The natural frequency of the tube bundle is an important parameter in FIV analysis. The heat transfer tubes of the spiral tube steam generator are made of spiral winding, and the positions between adjacent tubes are complex and variable. The winding direction between adjacent layers of the spiral heat transfer tubes can be in the same or opposite direction, so their spatial structure is complex. This article simulates and calculates the natural frequency of spiral heat transfer tubes in air and lead bismuth, and studies the influence of geometric structural parameters on their natural frequency. Based on simplifying constraints at both ends of the spiral tube bundles to fixed support, the computational fluid dynamics (CFD) simulation was carried out to reveal the effect mechanism of support number, spiral radius and spiral rising angle, on natural frequency of spiral tube bundles in air. The results indicate that an increase in spiral radius leads to a decrease in natural frequency, while an increase in support number leads to a significant increase in natural frequency, and the spiral rise angle has little effect on the natural frequency. Therefore, we can improve the stability of the spiral tube bundles by increasing support number or reducing spiral radius. Through CFD simulation calculation, the influence mechanism of tube bundle structural parameters on the natural frequency of spiral tube bundles in liquid metal environments was studied. The increase of layer pitch ratio and the same layer pitch ratio led to an increase in natural frequency. The research results provide data reference for the prediction of the natural frequency of heat transfer tubes and the study of FIV of tube bundles, which are of great significance for avoiding the failure of heat transfer tubes.
Numerical Simulation of Turbulent Cross Flow Over Helical Tube Bundles With Supporting Structures
Authors: Ma Anxiang , Xie Xiaoyang , Zhao Houjian , Li Xiaowei , Wu Xinxin 
Published: November 2024

Abstract
Helical tube bundles are usually used for the steam generator of high temperature gas-cooled reactor pebble-bed module (HTR-PM). The helical tube bundle is composed of several layers of helically coiled tubes, which are fixed by the supporting structures for preventing flow induced vibration. In the helical tube bundle, there are ideal flow paths separated by different layers of helical tubes. The blocking area ratio (the ratio of tube area to total area) of each flow path is different due to the helical diameter of each layer tubes, which may result in non-uniform flow. The presence of supporting structures will also affect the blocking area ratio, which may further increase the flow non-uniformity. The cross flow over helical tube bundles is complicated and difficult to predict. In the current investigation, RANS with SST k-ω model is used to simulate cross flow over a helical tube bundle with supporting structures. The influence of the supporting structure on the flow uniformity is investigated. The computational domain consists of 5 layers and 5 rows of helically coiled tubes. The influence of the helix angle is neglected. Due to the symmetry of the structure, only 1/3 part in the circumferential direction of the helical tube bundle is simulated. The velocity at the gap region between different layers of tubes is 19.74 m/s, corresponding to a Reynolds number of 28, 745. The simulation results are averaged over a long time and full length in the circumferential direction. Time-averaged distributions of velocity and pressure coefficient at different positions are obtained and compared. Multiple symmetrical secondary flows form in the z-normal plane and the secondary flows near the supporting structures are small. The averaged streamwise velocities in different flow paths are non-uniform. The flow oblique direction is investigated at different positions. In the region near the supporting structures, the flow direction is almost along the streamwise direction.
Fluid-Structure Coupling Analysis of U-Type Tube Bundle in PWR Steam Generator
Authors: Jiang Yingying , Xia Hong , Sun Lanxin , Li Shuang , Zhang Jinming 
Published: November 2024

Abstract
The structural integrity and safety of a steam generator are susceptible to flow-induced vibration. However, due to the intricate internal structure and thermal dynamics of the pressurized water reactor steam generator. Investigating the flow field characteristics and flow-induced vibration of the U-type tube bundle in steam generator holds pragmatic significance. The investigation delves into the flow field characteristics on both the primary and secondary sides of the U-type tubes, as well as the flow-induced vibration response, employing a multiphase flow model and a multiphysics coupling method. The distribution patterns of fluid pressure, temperature, and vapor fraction within the secondary side of both primary and secondary flow fields were investigated. Additionally, the distribution patterns of thermal stress and deformation under steady flow conditions in U-type tubes were determined. The results indicated that the support plates in the 8th and 9th layers have a significant impact on the thermal stress and deformation of U-type tubes. The temperature field and resultant thermal stress induced by fluid heat transfer on both sides of the U-type tube emerge as crucial factors affecting tube deformation.
Implementation of Domain Decomposition Parallel and gCMFD in MOCP: Method of Characteristics for Pebble-Bed
Authors: Hao Chen , Wen Yu Chen , Wang Yi Zhen 
Published: November 2024

Abstract
MOCP code is developed with Three-dimensional Method Of Characteristic (3D-MOC) for pebble-bed High Temperature Gas-cooled Reactor (HTGR) high-fidelity neutronics calculation. To further enhance its efficiency, domain decomposition parallel and gCMFD method are implemented in MOCP as alternatives to its previous calculation scheme, namely track grouping parallel and Macro-Track Transport Acceleration (MTTA) method. This paper provides parallel implementation details in track tracing, MOC sweeping and gCMFD solving. Parallel load balance is ensured by balancing subdomain volumes in each processor. Notably, Constructive Solid Geometry (CSG) based fine-coarse meshes mapping procedures are highlighted in this work. A simplified cylindrical model and HTR-10 whole-core model are used to demonstrate the performance of newly developed MOCP. It shows that MOCP parallel efficiency is around 80% using 160 MPI processors in the simplified cylindrical model. In HTR-10 whole-core model, the implemented domain decomposition parallel and gCMFD in newly developed MOCP could reduce around 70% core-hours compared with that using previous calculation scheme.
Experimental Study on the Sealing Performance of Core Barrel Flange for Swimming Pool-Type Low-Temperature Heating Reactor
Authors: Yu Song , Zhai Weiming , Cheng Daoxi , Zhou Ping , Ma Xiaoyao , Li Ruizhi , Xing Mingdi , Gao Weilong 
Published: November 2024

Abstract
The core barrel of the swimming pool-type low-temperature heating reactor is completely sealed by the structure of the flange itself. The failure of flange sealing will affect the thermal efficiency and safety of the reactor. In order to study the sealing performance of the core barrel flange, an experimental facility for measuring the leakage rate of core barrel flange was established by our team. The studies include measurements on the sealing capacity of different flange structure and factors affecting the leakage rate of flange. It was found that the leakage rate of the convex flange was smaller than that of the concave flange and the flat flange under the same experimental conditions. Thus, the convex flange is an optimal structure for the core barrel flange. For the core barrel flange with the same structure, the leakage rate increases with the flange surface roughness or internal-external pressure difference. To reduce the leakage rate of the flange sealing surface, the flange surface roughness of Ra3.2 could be selected. Furthermore, the leakage rate depends on the flange surface pressure and the fluid temperature. The experiments show that the maximum leakage rate of the core barrel flange is about 0.0047kg/s.
Influence of Nuclear Power Plant Interface Complexity on Operator Performance: A Modeling Study
Authors: Chen Pu , Tong Jiejuan 
Published: November 2024

Abstract
The main control room of the nuclear power plant is a complex human-machine tightly coupled system with frequent human-machine interactions, resulting in a high percentage of accidents/incidents due to operator factors. High temperature gas-cooled reactors adopt a multi-module design, with multiple reactor modules driving a single turbine. As a result, future high temperature gas cooled reactor - pebble-bed module nuclear power plants (HTR-PM) may require a single operator to monitor two or more nuclear steam supply system (NSSS), which may introduce new human factor safety issues. The rationality and ease of use of the human-machine interface (HMI) can reduce the cognitive load and the probability of error of the operator and ensure the safe operation of the nuclear power plant. In this paper, we conducted experiments on the HTR-PM600 simulator to simulate the main control room operator’s patrol task on two types of interfaces. And we simulate the cognitive behavior of operators in the above tasks based on the queuing network-model human processor (QN-MHP) method. The results of the simulations and experiments show that the QN-MHP method can well model the cognitive behavior of the nuclear power plant main control room operator. Moreover, the optimized HMI significantly reduces the operator’s task completion time and cognitive load, which proves that the QN-MHP method can support the optimization of HMI in nuclear power plants.
Design and Evaluation of Mitigation Measures for Tsinghua High Flux Reactor Loss-of-Coolant Accidents
Authors: Wang Zhuang , Xie Heng , Zhu Gan , Yeh Szu-En 
Published: November 2024

Abstract
Tsinghua High Flux Reactor is a planned new high flux research reactor in China. This paper designs a set of mitigation measures for loss-of-coolant accidents and preliminarily proposes the emergency core cooling scheme for Tsinghua High Flux Reactor. To verify the effectiveness of the current scheme, the corresponding Relap5 model is established, based on which break accidents at four different locations are simulated. The simulation results are presented. The core pressure, mass flow, coolant temperature, etc. in the transient process are studied and analyzed. According to the characteristics of the accidents, all break accidents are divided into five stages: non-shutdown stage, high-injection stage, low-injection stage, stable forced circulation stage and natural circulation stage. The non-shutdown stage is the most dangerous, and thus high core pressure and flow must be guaranteed. Sudo CHF correlations are used to predict critical heat flux. Results show that the minimum departure nucleate boiling ratios of all break accidents are larger than the safety limit of 1.5. Therefore, the current scheme can ensure core safety even if a double-ended rupture accident occurs.
A Review of the Performance Analysis of Plate Type Fuel Elements in High-Flux Reactors
Authors: Song Yue , Xie Heng , Lv Meng 
Published: November 2024

Abstract
High flux reactors, due to the characteristic of high neutron flux, are widely used in diverse areas such as research, development, education, and training. The plate type fuel elements need to be operated in harsh environments and consider the influence of high temperature, high pressure, intense radiation, and rapid coolant flow. These interactions among various physical fields have implications for the thermal, mechanical, material and chemical properties of the plate type fuel elements, and ultimately influence the high flux reactor’s safety performances. Therefore, to ensure the operational safety of high flux reactors, it is important to conduct safety performance analysis of plate type fuel elements under the effect of various physical fields during the design process of fuel elements in high flux reactors.
This paper studies the comprehensive analysis of domestic and international research results on plate type fuel elements and tries to conclude the impact of various factors on the operational performance of plate type elements, such as hydraulic effects, corrosion effects, thermal effects and irradiation effects. The analysis result will provide some references to further design and construction of high flux reactor.