bluuebunny/crossref_metadata_2025_split huggingface dataset:
Assessment and Analysis of Various Mechanisms in the Coalescence and Breakup Models for Upward Bubbly Flow
Authors: Guan Shunran , Han Jinyu , Zhao Chenru , Bo Hanliang 
Published: October 2021

Abstract
Accurate three-dimensional CFD simulation is of great importance to the flow and heat transfer of flow boiling in the steam generator in nuclear power plant, in which bubbly flow is the main flow pattern. One of the biggest challenges for the CFD simulation is the coalescence and breakup model for bubbles dispersed in the continuous fluid. Recently, several new bubble coalescence and breakup models have been developed combined with the inhomogeneous Multi-Size-Group (MUSIG) model in the Eulerian framework in addition to the standard closure model (Luo and Svendsen (1996), and Prince and Blanch (1990)), which has been widely adopted in the CFD simulations in bubbly flow using the Eulerian method. For example, Liao had developed a comprehensive closure model (2015), in which five collision mechanisms and four breakup mechanisms have been included in the model. However, proper coefficients for the Liao model must be determined to properly address the proportion for various mechanism in the coalescence and breakup models. In this paper, we compared the predictions using the standard closure model and the Liao model (2015) combined with the iMUSIG model for the case of adiabatic upward bubbly flow in vertical pipe, as well as the corresponding experimental results reported by Lucas et al (2008) which were obtained at TOPFLOW test facilities. Six simulated working conditions are chosen. The significance of various mechanisms on the bubble coalescence and breakup were analyzed based on the bubble size distributions in various heights obtained in the CFD predictions and experiments. Finally, a set of proper coefficients for various mechanisms in the coalescence and breakup models of Liao (2015) was developed for the simulated conditions.
Transfer Function Development and Dynamic Analysis of a Heat Pipe Cooled Reactor
Authors: Pu Songmao , Sun Peiwei , Wei Xinyu 
Published: October 2021

Abstract
The heat pipe cooled reactor adopts the solid-state reactor design concept and the heat is passively transferred out of the core through heat pipes. It is characterized by high inherent safety and simple operation and has broad application prospects in deep space exploration and propulsion, sea submarine navigation and exploration. The design of heat pipe cooled reactor is unique, and its dynamics are different from traditional water-cooled reactors. Therefore, it is necessary to develop its dynamic model and perform dynamic analysis, and in this paper, the study object of the heat pipe cooled reactor is the 100kW nuclear silent thermoelectric reactor (NUSTER-100).
A nonlinear dynamic model is derived from the conservation equations of mass, energy and momentum. Point reactor kinetics equations are adopted. The linear dynamic model is constructed by linearization of the nonlinear model based on the disturbance theory and the transfer function is further derived applying Laplace transform. Both models including the nonlinear model and transfer function model are established on the MATLAB & Simulink simulation platform.
Dynamic characteristic analysis contains time domain analysis and frequency domain analysis. For the time domain analysis, by introducing a variety of boundary condition disturbances, the results were compared with those from transfer function. The results are consistent and can correctly reflect the dynamic characteristics of the heat pipe cooled reactor. Therefore, the transfer function model can be applied to the subsequent design of the heat pipe cooled reactor power control system.
For the dynamic analysis, it is divided into time domain and frequency domain. The time domain is to observe the change of core power and sodium temperature by introducing reactivity disturbance. For the frequency domain, after drawing the Bode plot of the transfer function, the system’s characteristics at different frequencies are analyzed. In addition, it can provide a theoretical basis for the design of the heat pipe cooled reactor power control system.
Long-Term Simulation of Sodium Dynamics During a Large Leakage Sodium-Water Reaction
Authors: Bai Xi , Sun Peiwei , Luo Gang , Cao Huasong 
Published: October 2021

Abstract
The steam generator of the sodium-cooled reactor is the barrier between the secondary and third loops. When the heat transfer pipe breaks, the water/steam will pour into the sodium and a sodium-water reaction will occur. The pressure in the secondary loop will increase and the rupture disks will burst to reduce the pressure. For the safety analysis, the maximum pressure is limited. For the long term, the pressure pulse weakens and the fluid flow tends to a quasi-steady state several seconds or minutes after leak initiation. Therefore, it is necessary to develop a model to investigate the dynamics of the secondary loop in the long term.
The continuity equation, the momentum equation and the energy equation are used to derive the sodium flow model in the secondary loop. The sodium pressure and velocity are described by the one-dimensional differential equation. The lumped parameter method is applied and the differential equations are solved by the Euler method. FORTRAN language is compiled to develop the code. Critical equipment, including the steam generator, buffer tank, pump, rupture disks, and accident discharge tank are considered in the secondary loop model. The sodium velocity and pressure responses varying with time can be obtained. Compared with the data from the safety analysis report, the tendency of the sodium velocity and pressure is consistent. It is proved that the model is reasonable and effective to simulate and analyze the actual long-term effects of sodium dynamics.
An Original Distributed Simulation Method Applied to the Advanced Nuclear Power Plant Control Technology Hardware-in-the-Loop Simulation Verification Platform
Authors: Bowen Li , Zhe Dong , Di Jiang 
Published: October 2021

Abstract
Hardware-in-the-loop (HIL) simulation technology, where the part of a system to be verified adopts real objects, is one of the important methods for the research of advanced nuclear power plant (NPP) instrumentation and control (I&C) technology. With the development of advanced NPP I&C technology, especially the multi-module NPP technology, the HIL simulation technology is facing the challenge of communication signals booming and model extension to deal with the requirement of modules increasing and thermal-electricity generation. Driven by the above requirement of research and engineering, it is necessary to develop a novel HIL simulation technology that has well flexible scalability and avoids the high computational burden of the distributed control system (DCS). In this paper, an original distributed simulation method applied to the transformation extension of the NPP I&C HIL simulation verification platform is proposed. The initial opinion of the method is deploying a third-party system utilized for numerical simulation and form a close loop with DCS by network communication. With the support of third-party equipment represented by the real-time target machine, the functions of the system can be flexibly expanded through the MODBUS series protocol, and algorithms with high sampling frequency requirements can be deployed. The method has the characteristics of economical communication consumption, standardized and reliable communication protocol, and flexible downloading models and algorithms mean. Aside from this, due to the relative independence from DCS, the distribute simulation method is promising to be an original platform for verifying the technology advanced control or fault diagnosis in addition to DCS computing servers.
Application of Bayesian Classifiers for the Accident Diagnosis in Nuclear Power Plants
Authors: Qi Ben , Liang Jingang , Zhang Liguo , Tong Jiejuan , Yan Shu 
Published: October 2021

Abstract
Early identification of operational failures or faults is of vital importance for the prevention and mitigation of nuclear power plant (NPP) accidents. Various research efforts have been devoted to developing efficient accident diagnostic methods. Among them is the most promising artificial intelligence (AI) technology which makes use of plant status monitoring data for accident identification. However, the AI-based approaches are concerned with the difficulty in combining robustness and interpretability, i.e. the explanations of the features involved in modeling. This paper explores new accident diagnostic methods that are both robust and interpretable based on Bayesian classification algorithms. Specifically, three Bayesian classifiers including discrete naive Bayes classifier, Gaussian naive Bayes classifier, and Bayesian network are implemented and tested with simulation data for a PWR nuclear power plant (NPP) which consists of multiple types of the incident (LOCA, MSLB, SGTB) and dozens of monitoring parameters. The diagnostic accuracy and efficiency of these classifiers are inspected and compared quantificationally for all accident scenarios. The results show the Gaussian Naive Bayes method has the best accuracy which is close to 100%, followed by discrete Naive Bayes (above 97%), while the knowledge-based Bayesian network reveals plain accuracy due to limited usage of simulation data. The Naive Bayes classifiers also outperform Bayesian Network in diagnostic efficiency as they have simpler network structures. The Gaussian Naive Bayes even simplifies the implementation with straightforward data pre-processing.
Simulation for Predicting Condition of Plant Equipment
Authors: Kashiwase Shoichi , Osaki Kenji , Hatakeyama Makoto , Kaneko Tomokazu 
Published: October 2021

Abstract
Condition-based maintenance (CBM) is a maintenance program that recommends maintenance decisions based on maintenance information by condition monitoring. To apply CBM to industrial systems, it is important to diagnose condition of equipment appropriately based on the knowledge of abnormal condition of the equipment. In nuclear power plant, however, it is quite difficult to obtain the knowledge of abnormal condition. For this reason, simulation is expected to be an effective approach to estimate conditions of equipment instead of data acquisition. In this study, we focused on and built models for a motor operated valve and a pump, since the inspection cost is generally expensive due to the number of them in nuclear power plant. We considered a pump system and analyzed the measurement values in normal and abnormal conditions. Our analytical result reproduces the behavior of the measurement values and shows the characteristic deviation from normal condition in abnormal conditions.
Simulation Analysis and Optimization of Lubricating Oil System
Authors: Wu Qiongxiao , Wang Jianjun , Chen Jingming , Li Pengzheng 
Published: October 2021

Abstract
Based on the one-dimensional simulation model of lubricating oil system is established and analyzed by using FLOWMASTER software, this paper proposes a new method of optimizing lubricating oil system by PID technology. Ensure that the configuration requirements and control strategies of the relevant accessories of the simulation model are satisfied with the design requirements. Firstly, by simulating lubricating oil pressure fluctuation and lubricating oil flow distribution under Open/Close Valve in different opening and closing time, the optimal opening/closing time of Open/Close Valve is determined to be 0.2 s and 0.5 s respectively. Secondly, by writing the controller script file combined with a controller to realize automatic unloading relief valve simulation, determine the relief valve pressure regulating range of 0∼0.38 MPa, For precision of constant pressure valve of oil spill, the simulation results show that the average 10 m3/h flow caused by pressure changes of about 0.06 MPa. Under the flow sudden change signal of about 40 m3/h, the maximum pressure change is less than 0.1 MPa. Through the simulation results, it is found that most of the lubrication parts in the original design have the phenomenon of flow redundancy, which causes unnecessary pump power loss. The system is optimized by PID technology. By comparing the simulation results before and after optimization, it is found that the speed of constant displacement pump could be changed in time by PID controller, and the flow redundancy could be improved significantly, so the lubricating oil system could be lower consumption and achieve the purpose of optimization.
Liquid Film Behavior of Bottoming Liquid Jet in a Shallow Pool Measured by 3D-LIF
Authors: Yamamura Sota , Yoshida Hiroyuki , Horiguchi Naoki , Kaneko Akiko , Abe Yutaka 
Published: October 2021

Abstract
When the core meltdown accident occurs in the nuclear plant, molten corium falls into a coolant pool of the lower plenum. It is considered that the molten corium jet is broken up, cooled, and solidified with fuel-coolant interaction (FCI). However, the coolant pool could be a shallow condition by the leakage and evaporation of the coolant. In this situation, it is considered that the corium jet bottoms and spreads without the jet breakup. From the viewpoint of safety, understanding a jet behavior and estimating a cooling behavior are needed.
The purpose of this study is to clarify the mechanism of the liquid jet behavior in a shallow pool as the fundamental process for estimating the cooling behavior in the real machine. In this paper, we discuss the spreading behavior of the liquid jet after bottoming. The jet injection experiment was conducted using test fluids. By using the 3D-LIF method, the 3D visualization of the liquid jet was Successfully implemented. From the visualization result, the following behaviors were seen. After bottoming, the jet spread radially with the liquid film. As the jet spreading behavior, the liquid film was rolled up to the inside, and the vortex was formed. After a certain time, the vortex was broken. Then the flow and the number density of the fragment were changed.
Quantitative Measurements of Bubbles and Foam Flow Generated From Two-Phase Subcooled Flow Boiling of Seawater in a Vertical Annulus
Authors: Li Yuanjie , Shah Syed Waqar Ali , Pan Chin 
Published: October 2021

Abstract
Bubbly flow or foam flow could prevail in the subcooled flow boiling of seawater in an annulus. Compared with the bubbly flow of de-ionized water in the subcooled flow boiling, the bubble size of seawater is smaller. Further, bubbly flow with small bubbles or foam flow in a subcooled flow boiling of seawater would be affected by different heat fluxes, mass fluxes and inlet subcooling. The change in the boiling behaviors due to different conditions can be reflected in terms of variations of bubble characteristics. However, details of the characteristics of bubbles in seawater generated by the two-phase subcooled flow boiling are still limited. In the present study, bubble characteristics are measured and detailed by an image analytic method in the software Image J. The subcooled flow boiling experiments with different heat fluxes are conducted in a vertical upward annulus with artificial seawater at the inlet temperature of 90°C and mass flux of 1200 kg/m2s. The two-phase flow patterns are recorded by the high-speed camera in 6000 frame/s. To conduct measurements of bubble characteristics using Image J, images are selected as one set of measurement from the videos recording under different heat fluxes. After preprocessing of images, measurements of bubbles and fitting procedures with ellipses are conducted. Averaged values of descriptors of bubbles are obtained, including areas, perimeters, lengths of major axes, lengths of minor axes, circularities, angles, solidities, and aspect ratios, etc. It provides a possible way to investigate the details of bubbly flow or foam flow in the subcooled flow boiling of seawater.
Characteristics of Two-Phase Flow in Packed Bed Systems
Authors: Yasugi Noriaki , Fujitsu Akito , Odaira Naoya , Ito Daisuke , Ito Kei , Saito Yasushi 
Published: October 2021

Abstract
Two-phase pressure drop in the debris has been studied by many researchers in relation to the debris cooling characteristics during a severe accident in a nuclear reactor. However, its flow regime transition of the two-phase flow in the debris has not been well understood, which strongly affects the interfacial drag and the pressure drop. Conventional models for gas-liquid two-phase flow pressure drop have not been established well to evaluate interfacial drag accurately. In this study, high-speed imaging of a two-dimensional network model was performed to clarify the effect of flow pattern on interfacial drag and pressure drop. Normally it would be very difficult to visualize such two-phase flow behavior in an ordinary packed bed due to the reflection/refraction of light and/or overlapping bubbles, even if the test section is made of transparent materials. Therefore, in this study, a test section, which simulates two-dimensional network of porous structures, was fabricated to avoid the overlapping bubbles. By using a high-speed imaging of the two-dimensional network model, two-phase flow pattern in the porous structure have been identified. From the experimental results, it was suggested that the interfacial drag term should be modified in the gas-liquid two-phase flow pressure drop model.
Tritium Transport Modeling and Analysis for HCCB Blanket of CFETR
Authors: Zhang Baorui , Xia Zhaoyang , Zhou Zhiwei 
Published: October 2021

Abstract
China Fusion Engineering Test Reactor (CFETR) is aimed to meet the technical gaps between International Thermonuclear Experimental Reactor (ITER) and future fusion DEMO reactors, tritium self-sufficiency must be guaranteed for the operation of fusion reactors. Helium Cooled Ceramic Breeder (HCCB) blanket is one of the promising candidate blanket schemes for CFETR. Tritium generated in Li4SiO4 pebble bed is carried out by purge gas, it is crucial to study the tritium behavior in the blanket from the tritium fuel cycle point of view. In this paper, the effective diffusivity of tritium in purge gas is conducted based on Discrete Element Method (DEM) and Computational Fluid Dynamics (CFD) coupled simulation considering the influence of the pebble bed structure. The flow characters analysis of purge gas is carried out by CFD simulation as well, which includes velocity field distribution, pressure field distribution and pressure drop. Based on the macro parameters conducted above, the tritium transport process in blanket including tritium permeation into coolant, tritium inventory in blanket and tritium carried out by purge gas is simulated based on a 2D model set up in COMSOL Multiphysics, the results can be the reference for the design of Tritium Extraction System (TES) and Coolant Purification System (CPS).
Estimation of the Amount of I-129 in the Environment Generated Due to the Decay of Te-129m Discharged by the Fukushima NPS Accident
Authors: Sato Haruo 
Published: October 2021

Abstract
The accident at the Fukushima Daiichi Nuclear Power Station (1F-NPS) occurred following the Great East Japan Earthquake in March 2011, and led to the release of volatile radionuclides, which were deposited on the environment such as soils, forests, residential land, etc. in the Fukushima prefecture and the neighbouring areas. From immediately after the accident, air dose rate monitoring and analyses of radionuclides in soil, etc. were conducted in a wide range of area. Consequently, volatile radionuclides such as I-131, Cs-134, Cs-137, Te-129m, and Ag-110m were detected in a wide range of area, where was distant over 60km from 1F-NPS. Since I-131 is short-lived, it approximately disappeared after a few months. After the decay of short-lived I-131, Radiocaesium (Cs-134 and Cs-137) are the main contributors to radiation dose rate.
On the other hand, Te-129m, of which amount was approximately the same level as both Radiocaesium, was also detected although its half-life is short about 1 month. Since very long-lived I-129 is generated due to the decay of Te-129m, the accumulation of I-129 in the environment by accident is worried. Te-129m decays to I-129 by the following 2 decay series; Te-129m → Te-129 (half-life: 69.6min) → I-129 (64%), Te-129m → I-129 (36%). Since the half-lives of Te-129m and Te-129 are 33.6d and 69.6min, respectively, radioactive equilibrium is established between both nuclides, but it is not established between Te-129 and I-129 and between Te-129m and I-129. In this case, the amount of I-129 increases due to the decay of Te-129m, but also the maximum value of the amount theoretically exists.
In this study, the amount of I-129 in the environment generated due to the decay of Te-129m discharged by the 1F-NPS accident was estimated based on the analysed data of Te-129m in soil. The analysis was conducted for 3 investigation locations (Okuma Town, ca. 2.7km from 1F-NPS, Naraha Town, ca. 19km from 1F-NPS, and Fukushima City, ca. 60km from 1F-NPS) among the analysed data (as of 14 June 2011) of soil samples obtained at about 2,200 investigation locations in the Fukushima prefecture and the neighbouring prefectures. The maximum amount of I-129 was estimated 1.03 × 10−1 (Bq/m2) after about 1,000d even for the maximum deposition amount of Te-129m (2.7 (MBq/m2) as of 14 June 2011). This is approximately the same level as radioactivity concentration in the environment (5.75 × 10−3 ∼ 2.96 × 10−1 (Bq/m2)). Additionally, the results of field investigation around 1F-NPS also support this conclusion.
Comparison on HAPPY200 Reactor With Different Type Fuel Assembly
Authors: Sun Canhui , Zhou Tao , Chen Yaodong , Meng Zhaocan 
Published: October 2021

Abstract
In order to meet the new challenges on the energy and environment, a new conceptual design of nuclear heating reactor (HAPPY200) is developed. HAPPY200 is a two-loop low pressure water reactor with low thermal parameters and passive safety systems. The outlet temperature is 120 °C, the inlet temperature is 80 °C, and the operating pressure is 0.6MPa. In this study, the thermal performance problems of big and small fuel assemblies are discussed for HAPPY200. Under the same total power condition, the small assembly core is better considering thermal performance. And under the same average rod power condition, the small assembly core condition 2 is better considering thermal performance. Considering maturity and economic factor comprehensively, the big assembly core is chosen in the near term.
Research on the Air-Water Flow Regime and Characteristics in Rectangular Channel
Authors: He Qingche , Pan Liangming , Zhang Luteng , Yan Meiyue , Xu Wangtao 
Published: October 2021

Abstract
Two-phase Flow is widely involved in reactor design and is directly relevant to reactor safety. However, the flow regime in narrow rectangular channels still needs further study because it has a considerable difference from tube and bundle channels. To investigate the two-phase flow regime and interfacial structure characteristics, the air-water experiment with an adiabatic vertical channel of 4 × 66 × 1800, 6 × 66 × 1800 mm have been conducted under atmosphere pressure condition. The impedance void meter was used to measure the global void fraction in narrow rectangular channels. A high-speed camera was used to record the profiles of the flow regime. The flow regime was identified by the random forest clustering algorithm based on a training sample. The profiles of different parameters, including void fraction, pressure loss at Z/D = 150, were analyzed in this paper. Furthermore, based on the parameters’ distribution, the regime transition criteria in narrow rectangular channels were discussed. It is shown that the transition from bubble to slug flow always occurred when the average void fraction is 0.17–0.2. The transition value is 0.57–0.62 when the slug Flow changes to the churn-turbulent Flow and 0.78–0.8 from churn-turbulent to annular Flow. The constant used in the Lockhart-Martinelli correlation is found to calculate the frictional pressure drop in a rectangular channel. Furthermore, the drift-model applied to the rectangular channel is verified.