bluuebunny/crossref_metadata_2025_split huggingface dataset:
Study on Measures During Loss of Normal Feedwater Accident for AP1000 NPP in Lower Power Operation
Authors: Ma Baisong , Miao Zhengqiang , Zhuang Yaping , Hou Weili 
Published: October 2020

Abstract
If an accident occurred in a plant in lower power operation, a series of problems could be caused by low residual heat and excessive removal capability of heat sinks. Taking a real accident as example, this paper discussed how to deal with loss of normal feedwater in AP1000 NPP during lower power operation. The level of Steam Generators (SG) will be rapidly reduced in case of Main Feedwater loss, and Startup Feedwater pump will start. If the reactor is not shut down immediately, the Startup Feedwater will not be enough to prevent SG level from falling, and automatic reactor trip will be inevitable. Therefore, in event of loss of Main Feedwater which cannot be recovered instantly, the reactor should be shut down manually to maintain SG level as high as possible. Appropriate measures should be taken to avoid excessive cooling after trip. Attention should be paid to SG level and feedwater flow to avoid Passive Residual Heat Removal Heat Exchanger (PRHR HX) actuation during regulating Startup Feedwater. Otherwise, PRHR HX may be actuated unexpectedly. With Reactor Coolant Pumps operating, PRHR HX heat removal capacity is strong. The Cold Leg temperature will drop rapidly because of PRHR HX and other heat sinks. In this case, the most effective measure to stabilize temperature of Cold legs is to isolate Streamline. Otherwise, structure integrity of the Reactor Pressure Vessel (RPV) will be challenged under Pressurized Thermal Shock (PTS) due to rapid drop of Cold leg temperature. In order to alleviate PTS, depressurization shall be performed and further cooling shall be suspended. Since PRHR HX is only connected to Loop 1, coolant in Hot Leg of Loop 2 is still kept at high temperature. Therefore, necessary measures shall be taken to prevent coolant flash in Loop 2 during the process of depressurization, which may fill the Pressurizer up with water. The RPV Head Vent valves perform safety-related function by preventing Pressurizer overfill in certain design basis events in AP1000 NPP. However, emergency letdown by opening the valves cannot effectively reduce Pressurizer level when coolant flashed in Loop 2. Because of mitigating PTS, opening RPV Head Vent valves, and steam condensation in Loop 2, the subcooling of the coolant at the core outlet is likely to drop below 6 °C or even 0 °C.
Study on the Effect of Different Factors of Displacement Cascades in Alpha-Fe by Molecular Dynamics Simulations
Authors: Lin Pandong , Nie Junfeng , Liu Meidan 
Published: October 2020

Abstract
As the key component of RPV steel, α-Fe is under neutron irradiation during its long-term service, and lattice atoms of α-Fe are knocked by neutrons, which leads to irradiation damage. In this paper, molecular dynamics method is conducted to investigate the effect of temperature, vacancy concentration and tensile strain on irradiation-induced damage by displacement cascade simulations in α-Fe. The simulations are performed with primary knock-on atom energies ranging from 0.1 to 5 keV, temperature ranging from 100 to 500K, vacancy concentration ranging from 0% to 1% and applied tensile strain ranging from 0 to 5%. The time evolution of defects produced during displacement cascade can be obtained where the surviving number of Frenkel pairs increases rapidly at first, then decrease and comes to stability finally. The influence of these factors on defect production can be concluded as following: The increase of PKA energy, vacancy concentration and applied tensile strain can lead to the increase of defect numbers. In contrast, the increase of temperature decreases the defect numbers. Vacancies and interstitials cluster size distributions are varied in different case. The results are meaningful to describe some microcosmic mechanisms of RPV steels in nuclear system.
Multi-Group Staggered Tests for Highly Redundant Systems
Authors: Soga Shota 
Published: October 2020

Abstract
A reactor protection system (RPS) in a boiling water reactor (BWR) is a unique system since it represents high redundancy compared with other safety-related systems in the BWR. Owning to its high redundancy, NUREG-5500 vol.3 showed that unreliability of the General Electric BWR4-type RPS is dominated by common-cause failures (CCFs) in its highly redundant components. Since the staggered test scheme can reduce the period of latent CCF states of redundant standby components, the scheme is a good candidate to improve the reliability of the RPS. However, the scheme is impracticable for the RPS because of its high redundancy.
In this paper, a new concept of a “multi-group staggered test scheme” for a highly redundant system is proposed. In the multi-group staggered test scheme, components are arranged into several groups, and all components in a group are tested simultaneously. This paper presents a preliminary example of the new scheme that is applied to control rod notch tests, in which all withdrawn control rods are moved at least one notch. It is shown that the new scheme can reduce the CCF risk in RPS.
Experimental Investigation of Particle Decontamination Efficiency in a Single-Bubble by Pool Scrubbing
Authors: Fujiwara Kota , Nakamura Yuki , Yoshida Kohei , Kaneko Akiko , Abe Yutaka 
Published: October 2020

Abstract
Nuclear power plant (NPP) safety has become a public issue since the Fukushima daiichi NPP accident. In order to evaluate the risks caused by severe accidents (SAs), it is very important to understand the on-site source term events. One of the important unsolved source term events is the decontamination efficiency of fission products (FPs) in the suppression chamber by pool scrubbing. Therefore, a mechanistic model to analyze the particle decontamination efficiency by pool scrubbing is highly regarded. Despite the demand, particle decontamination mechanism by pool scrubbing has never been understood due to the complexity of phenomena.
In our experiment, we aim to develop a reliable mechanistic model to evaluate particle decontamination efficiency of pool scrubbing by conducting separate effect tests. As to obtain the fundamental process of particle decontamination from gas to liquid-phase, we focused on decontamination factor (DF) of particle from a single bubble. However, it is very difficult to calculate the initial particle concentration inside the bubble. Therefore, in our experiment, we developed a method to measure the internal particle concentration inside the bubble by combining image processing and particle measurement.
By using the experimental results, we succeeded to obtain reasonable DF for glycerin particles and CsI particles as a simulant particle for FPs. From the experimental results, detailed particle decontamination efficiency for various submergence were measured. The results tend show that DF increase linearly as submergence increases which suggests that DF is constant on bubble rise region. Moreover, the fact that glycerin particle with larger particle diameter takes a higher value shows that particle diameter significantly affects DF.
Review of Application on Dynamic Fault Tree Method in Nuclear Power Plants
Authors: Guohua Wu , Diping Yuan , Yiqing Xiao , Jiaxin Wang 
Published: October 2020

Abstract
Fault tree analysis (FTA) is one of the most important methods of probabilistic risk assessment (PRA). The fault state of the system is taken. While traditional FTA is based on static failure model. FTA is not applicable for systems that include redundant, sequence-related systems. At the same time, nuclear power plants (NPPs) contains a large number of redundant equipment, and FTA is difficult to solve these dynamic problems. Therefore, it is necessary to use dynamic fault tree analysis (DFTA) for PRA. In DFTA research, the modular analysis method was first proposed. The modular method divides the dynamic fault tree into a dynamic fault tree and a static fault tree. Among them, the dynamic fault tree is analyzed using a Markov chain, and the static fault tree is studied using a binary decision diagrams method. However, the shortcomings are that when the system is complicated, the information explosion in the Markov chain is appeared. To solve this problem, a dynamic fault tree is transformed into a Bayesian network. At the same time, to verify the feasibility of the method, Monte Carlo random sampling was used to evaluate the method. Other methods are relatively infrequently studied.
In this paper, firstly, status of dynamic fault trees has been investigated. Secondly, each method is analyzed and the problems of dynamic fault tree are described. Finally, a survey and analysis on the dynamic fault tree is finished, and the main problems of the dynamic fault tree are: information explosion; the lack of commercial software to apply to engineering. Through this review, we hope to play a certain guiding role in the subsequent research on dynamic fault trees.
Development of a Flow Network Calculation Code (FNCC) for High Temperature Gas-Cooled Reactors (HTGRs)
Authors: Aoki Takeshi , Isaka Kazuyoshi , Sato Hiroyuki , Ohashi Hirofumi 
Published: October 2020

Abstract
A core of the prismatic high temperature gas-cooled reactor (HTGR) consists of fuel blocks, control rod guide blocks and replaceable reflector blocks with surrounding permanent reflector blocks. The core flowrate available for effective cooling of fuels, fuel blocks and control rod guide blocks has to be evaluated to confirm that maximum fuel temperature satisfies fuel design limits. In the evaluation, unintended flows such as gap flows between columns, cross flows between column layers and gap flows between permanent reflectors should be analyzed for the prismatic-type HTGR design. The flow distribution has to be evaluated preciously with minimizing the unintended flows.
In the HTGR design, flow distribution analysis has to take into account the interaction of the gaps between the graphite structures forming unintended flow paths and flow conditions such as coolant temperature and pressure by thermal and radiation deformations of the graphite structure.
In the present study, a user-friendly flow network calculation code (FNCC) has been developed on the basis of experiences of High Temperature engineering Test Reactor (HTTR) design for HTGR design with enhanced compatibility with other HTGR design codes and with considering graphite block deformation in iteration process without manual control. The code is able to solve the flow distribution based on conservation equations on the mass, momentum and energy by converging iteration calculations along with update of gap information in each calculation step. The validation result using FNCC provides good agreement with a result of the one-column flow distribution experiment.
Synergetic Oxidation in Alkaline In-Situ Leaching Uranium: A Preliminary Case Study
Authors: Liao Wensheng , Que Weimin , Wang Liming , Du Zhiming 
Published: October 2020

Abstract
In alkaline in-situ leaching uranium, oxygen is the most common oxidizer with bicarbonate as a complexing agent. For those sandstone uranium deposits with strongly reductive capacity or complicated hydrogeological environment, the oxidation by oxygen is low efficiency. An efficient leaching method, therefore, is needed for these uranium deposits. In this study, a typical sandstone uranium deposit which characterizes with high TDS and high chloride content in groundwater and intractable uranium leach is selected to investigate the effects of synergetic oxidation by a strong oxidant with oxygen. Based on the research on batch leach, pressure leach and field trials, the oxidants such as hydrogen peroxide, potassium permanganate and sodium dichloroisocyanurate (NaDCC) are tested. The results of pressure batch leach indicate that synergetic oxidization is achieved by NaDCC in oxygen leaching process. Leaching tests indicate that a minor oxidizer of NaDCC shows good synergetic oxidization with oxygen and leaching effects on uranium minerals. The results also demonstrate that hydrogen peroxide shows no oxidation effects when it is used as a single oxidant. While potassium permanganate shows good oxidation on uranium when it is used as a single oxidant, however, it leads inhibiting effects on oxygen oxidation on uranium minerals. The further field tests are conducted to study the synergetic effects of oxygen with and without sodium dichloroisocyanurate. The preliminary results indicate that a fast leach is observed by the composite oxidants in early stage while no synergetic leach is found after 200 days. Further studies should be conducted in laboratory experiments and pilot scale tests for its potential applications.
Numerical Study of Collision Behavior of Melt Drops During Fuel-Coolant Interaction
Authors: Wen Panpan , Li Gen , Gao Jinchen , Li Yupeng , Yamaji Akifumi , Yan Junjie 
Published: October 2020

Abstract
The collision dynamics between two droplets plays an important role in various disciplines of nature and practical interests, such as fuel-coolant interaction (FCI), fuel combustion in engines, and various spraying process. FCI presents in nuclear reactor severe accident when the melt relocates into the coolant in the lower head with violent disturbance and vigorous heat transfer. The purpose of this study is to investigate the collision behavior of melt droplets during fuel-coolant interaction. The collision of two equal-sized droplets has been simulated in 3D by using the volume of fluid (VOF) and adaptive mesh refinement method. The numerical simulations of tetradecane droplet collision were carried out to validate the numerical methods. The results showed good agreement with the experiments. Furthermore, the simulations of uranium dioxide (UO2) droplets collision in coolant were carried out. The results showed that the contact area between droplets and coolant increased with time first and then decreased. With the increase of Weber number, the contact area of maximum in the droplet collision increased. Break happened in the later period and many child droplets formed. The number of child droplets increased with the increase of Weber number. In addition, the size distribution of little droplets was investigated.
Proportional-Integral Disturbance Observer of Nuclear Reactors
Authors: Dong Zhe 
Published: October 2020

Abstract
A proportional-integral disturbance observer (PI-DO) for monitoring nuclear reactors is newly proposed, which is driven by the measurements of neutron flux and coolant temperature at reactor inlet as well as their integrations. This PI-DO provides a globally asymptotic estimation with a bounded steady-state error for the reactor key process variables as well as the total disturbances in channels of the neutron kinetics and primary coolant thermal-hydraulics. Moreover, the PI-DO is applied to reconstruct the unmeasurable state variables and total disturbances of a nuclear heating reactor (NHR). Numerical simulation results not only verify the theoretic analysis but also show both the satisfactory performance and the influence of observer parameters.
Experimental Study on Flow Patterns of Decaying Swirling Gas-Liquid Flow in a Horizontal Pipe
Authors: Liu Shuai , Liu Li , Zhang Jiarong , Gu Hanyang 
Published: October 2020

Abstract
Swirling flow is one of the well-recognized techniques to control the working process. This special flow is widely adopted in swirl vane separators in nuclear steam generator (SG) for water droplet separation and the fission gas removal system in Thorium Molten Salt Reactor (TMSR) for gas bubble separation. Since the parameters such as separation efficiency, pressure drop and mass and heat transfer rate are strongly dependent on the flow pattern, the accurate prediction of flow patterns and their transitions is extremely important for the proper design, operation and optimization of swirling two-phase flow systems. In this paper, using air and water as working fluids, a visualization experiment is carried out to study the gas-liquid flow in a horizontal pipe containing a swirler with four helical vanes. The test pipe is 5 m in length and 30 mm in diameter. Firstly, five typical flow patterns of swirling gas-liquid flow at the outlet of the swirler are classified and defined, these being spiral chain, swirling gas column, swirling intermittent, swirling annular and swirling ribbon flow. Being affected by the different gas and liquid flow rate of non-swirling flow, it is found that the same non-swirling flow can change into different swirling flow patterns. After that, the evolution of various swirling flow patterns along the streamwise direction is analyzed considering the influence of swirl attenuation. The results indicate that the same swirling flow pattern can transform into a variety of swirling flow patterns and subsequent non-swirling flow patterns. Finally, the flow pattern maps at different positions downstream of the swirler are presented.
Friction Factor Correlation for Molten Metals Flow in a 3-Rod Bundle in Cylindrical Shroud, With Smooth and Scalloped Inner Wall
Authors: Schriener Timothy , El-Genk Mohamed S. 
Published: October 2020

Abstract
Thermal-hydraulic experiments to help develop and qualify new fuel designs for advanced liquid metal cooled reactors could use reduced-size triangular pitch, 3 rods within a circular shroud. In such an arrangement, scalloping the shroud wall helps achieve lateral flow uniformity and high velocity, but at the expense of increasing the pressure losses. Owing to the practical challenges of conducting experimental investigations with molten lead, Computational Fluid Dynamics (CFD) could a practical alternative to generate large enough database to develop a friction factor correlation. This work performed CFD analyses of molten lead flow in 3-rod bundle with circular shroud to investigate the effect of scalloping the shroud wall on the lateral flow uniformity and pressure. The developed friction factor correlation is very smooth, spanning the laminar, transition, and turbulent flow regions. Scalloping the shroud wall increases the pressure losses, but not the friction factor in the transition and turbulent flow regions. In the laminar flow region, however, scalloping the shroud wall only slightly increases the friction factor.
Emulation Methodology of Programmable Logic Controllers for Cybersecurity Applications
Authors: Fasano Raymond , Lamb Christopher , El Genk Mohamed , Schriener Timothy , Hahn Andrew 
Published: October 2020

Abstract
A programmable logic controller (PLC) emulation methodology can dramatically reduce the cost of high-fidelity operational technology (OT) network emulation without compromising specific functionality. A PLC emulation methodology is developed as part of an ongoing effort at the University of New Mexico’s Institute for Space and Nuclear Power Studies (UNM-ISNPS) in collaboration with Sandia National Laboratories (SNL) to develop an emulytic™ platform to support cybersecurity analyses of the instrumentation and control (I&C) systems of pressurized water reactors (PWRs). This methodology identifies and characterizes key physical and digital signatures of interest. The obtained and displayed digital signatures include the network response, traffic, and software version, while the selected physical signatures include the actuation response time and sampling time. An extensive validation analysis is performed to characterize the signatures of the real, hardware-based PLC and the emulated PLC. These signatures are then compared to quantify differences and identify optimum settings for the emulation fidelity.
Evaluation of Irradiation Embrittlement of the Weld Metals in the Chinese RPV
Authors: Cao Yupeng , He Yinbiao , Huang Yifeng , Wang Binxi , Yu Yan , Li Hui , Hui Hu 
Published: October 2020

Abstract
Reactor pressure vessel (RPV) operates under high temperatures and pressures and is exposed to relatively high neutron radiation. RPV is considered to be irreplaceable, which is the most limiting factor for the lifetime of a nuclear power plant. As the most severe ageing degradation mechanism in RPV materials, irradiation embrittlement is a major issue affecting the integrity through the service life of a RPV. Our previous paper (ASME PVP2019-93615[1]) introduced our project for assessment of irradiation embrittlement of the materials for the Chinese RPVs to verify the 60-year design life, in which the specimens made of the RPV base material manufactured in China, the SA-508 Gr.3 Cl.1 forging, and the different types of weld metals were irradiated in the high fluence engineering test reactor (HFETR). The paper analyzed extent irradiation damage of the forging in terms of mechanical properties. As another part of the project, this paper concentrates on the evaluation of the weld metals in the same project. Tensile tests, Charpy impact tests and fracture toughness tests by master curve approach were carried out for the three types of weld metals subjected to different irradiation fluences (2.6E19n/cm2, 8.9E19n/cm2). Comparison of the mechanical properties of the irradiated and the unirradiated materials is made. The irradiation resistance of the weld metals in our project is also compared with the data in the literatures.
Numerical Investigation of the Obstacle on Hydrogen Distribution in a Vessel
Authors: Wang Tianlin , Tong Lili , Cao Xuewu 
Published: October 2020

Abstract
Hydrogen combustion and explosion is an important safety issue in nuclear power plants (NPPs) containment during postulated severe accidents or in utilization of hydrogen. It is significant to understand the hydrogen flow and distribution in space for mitigating hydrogen risk. In this paper, a numerical model to investigate hydrogen flow and distribution in a vessel is established using computational fluid dynamics (CFD) code GASFLOW. Hydrogen is simulated by helium which is used to study the hydrogen distribution. The k-ε turbulent model is selected to establish the numerical model, and the numerical model which has no obstacle inside the vessel or includes the obstacle inside is verified under medium momentum conditions of injected gas by comparing numerical results with experimental data. Regardless of the presence of the obstacle in the vessel, helium stratification occurs under all momentum condition of injected gas. When the obstacle is present, it blocks the flow path of the injected helium to the upper space, then the helium volume concentration in upper space is lower than the condition that there is no obstacle in the vessel. As the initial Froude number increases from 0.19 to 19.29, the flow mechanism around the obstacle under high momentum condition of injected gas is different from that under medium or low momentum conditions. Consequently, the boundary of the helium stratification moves down, and the distribution of helium looks more uniform in most area of the vessel for high momentum conditions of injected gas.