bluuebunny/crossref_metadata_2025_split huggingface dataset:
Understanding Ammonia/Hydrogen Fuel Combustion Modeling in a Quiescent Environment
Authors: Shaalan Amr , Nasim Md Nayer , Mack J. Hunter , Van Dam Noah , Assanis Dimitris 
Published: November 2022

Abstract
Ammonia and Hydrogen are attractive alternative fuels for a zero-carbon combustion solution that can rapidly decarbonize the transportation industry. Understanding the chemical behavior and combustion characteristics of these fuels individually, as well as blended together, is pivotal to ensuring their widespread adoption and utilization. Furthermore, in the era of computer-aided engineering, it is critical to evaluate our ability to computationally model the chemical reactivity of these two fuels and validate predictions of experimentally observed phenomena using multi-dimensional simulations. In this study, ammonia/hydrogen chemical kinetics mechanisms available from the research literature are investigated through 0-D, 1-D, and 3-D simulations. The 0-D and 1-D simulations were carried out to understand the ignition delay and laminar flame speeds, respectively, at different operating pressures and temperatures. 3-D simulations were also performed to test the fuels’ behavior in a closed volume combustion chamber. The multi-dimensional computational results were compared against optically measured experimental data available in recent publications. Specifically, a comparison of unstretched flame speeds determined from stretched flame speeds of post-processed computational results is made. Lean and rich combustion limits have been computationally evaluated as well. Lastly, observed physical buoyancy effects were reproducible in a quiescent computational environment leading to increased confidence in using the evaluated chemical kinetics mechanisms for high-fidelity reciprocating piston engine computational research and development.
Temperature Dependent Removal Efficiency of Crankcase Control Devices
Authors: Hicks Myles D. , Northrop William F. , Potratz Daniel , Kittelson David B. 
Published: November 2022

Abstract
Oil aerosols formed in the crankcase of reciprocating engines are a significant source of particulate matter and may lead to engine component degradation when used in closed crankcase ventilation (CV) systems or constitute a significant emission source in open CV systems. In modern CV systems, filtration or inertial separation is used to collect oil particles. These devices are highly efficient over a range of flow rates and temperatures; however, some system conditions result in a decrease in efficiency with increasing temperature in a manner not predicted by theory.
Experiments were performed on a bench system that introduced atomized oil particles into a clean, pre-heated, air stream that was conveyed isothermally through a heated duct to an oven containing the removal device, either a coalescing filter or an inertial separator, followed by another heated section and an exit duct. The aerosol was sampled using identical upstream and downstream isokinetic sampling systems and characterized using a Dekati Electrical Low-Pressure Impactor (ELPI) that measured particle concentration and size in the range from 0.043 to 8.46μm aerodynamic diameter.
Neither single fiber efficiency (SFE) filtration theory nor physical arguments describing the performance of the inertial separator predict a significant dependence total and fractional efficiency and most penetrating particle size (MPPS) on temperature in the range explored here, 25 to 115°C, but in many cases our measurements show MPPS shifting to smaller size and total efficiency decreasing with increasing temperature. These observations are consistent with the hypothesis that, at device temperature, the particles being processed are smaller than the particles being measured, at room temperature. Particles shrink by evaporation as they pass through the heated sections, but cool and grow by condensation as the sample cools in the sampling lines or in the exit duct.
Reported carbon number distributions of typical lubricating oils and evaporation kinetics indicate that droplets in the size range investigated may lose more than half of their mass under temperature conditions, up to 115°C, found in typical crankcase removal devices under engine high-load conditions. These droplets are in a dynamic balance with their vapors and may change during sampling and measurement. Great care must be taken in design of systems used to characterize devices intended to remove volatile droplets and interpretation of measurements made, as the size of the particles measured may not be the same as the size of particles passing through the device.
Complementary Simulations to Determine Heat Transfer Coefficients and the Maximum Heat Flux in Multi-Nozzle Spray Cooling Experiments
Authors: Du Yuxuan , Bandaru Satya V. Ravikumar , Villanueva Walter 
Published: November 2022

Abstract
For Light Water Reactor (LWR) safety, spray cooling during severe accidents is one of the promising approaches to achieve In-Vessel Retention of corium by External Reactor Vessel Cooling (IVR-ERVC). To study the efficiency of multi-nozzle spray cooling (nozzles of 2 × 3 matrix) on a downward-facing FeCrAl heated surface, a lab-scale experimental facility was built. It should be emphasized, however, that a direct measurement of Heat Transfer Coefficient (HTC) on the sprayed side is challenging due to the strong interference of water flow and intrusiveness of standard instrumentation methods. In this paper, a 3D numerical model has been established with the same geometric and material parameters as the foil sample in a multi-nozzle upward spray cooling. Given the experimental temperature profiles on the sample’s dry side measured by an IR camera, the complementary numerical simulations have revealed the HTCs and corresponding temperature profiles on the sprayed side, which enabled the prediction of the maximum heat fluxes (MHFs). The maximum heat fluxes for the given spray cooling conditions can reach up to 3.25MW/m2, which is more than adequate for what is required for a successful IVR-ERVC for high-power reactors. At the same time, the maximum temperature on the dry side at the highest input power is still much lower than the expected failure temperature of the sample material.
Effect of the Plasticity of Piping and Support Structures on the Seismic Response of Piping Systems
Authors: Okuda Takahiro , Takahashi Hideki , Watakabe Tomoyoshi , Morishita Masaki 
Published: November 2022

Abstract
This study describes the results of seismic analyses conducted to clarify the effect of the plastic deformation of a pipe and its support structures on the seismic response of the entire piping system. The first type of analyses includes the investigation of plasticity with regard to the following parameters: (1) none, (2) piping, (3) support structure, and (4) piping and support structure. The second type of analyses includes the parametric survey of the yield load and post-yield stiffness of the restoring force characteristics of support structures.
The results of the first type of analyses have demonstrated significant implications on the seismic design of piping systems based on the elastic-plastic response analysis. First, the piping response, i.e., elbow strain, is significantly reduced when plasticity is considered in the piping support. On the contrary, the piping support response is not so much affected by the piping plasticity.
The results of the second type of analyses showed that the yield load of the support structure significantly affects the response of the piping system, whereas the post-yield stiffness does not affect much compared with the yield load.
Numerical Simulation of Aerosol Generation and Aerosol Spray Scavenging During Fukushima Daiichi Decommissioning
Authors: Liang Hui , Nejdet Erkan , Wang Kai , Sharma Avadhesh Kumar , Suzuki Shunichi 
Published: November 2022

Abstract
In the future decommissioning plan of the damaged Fukushima Daiichi nuclear reactors, the melted and re-solidified fuel debris at the bottom of the reactor pressure vessel and primary containment vessel will be cut into small pieces before retrieving them from reactor buildings. Submicron aerosol particles that generated during cutting operations must be removed inside the containment vessel before they escape to the environment. The existed spray system inside the containment vessel can be used to remove these aerosol particles under different collection mechanisms. In this study, a new numerical model is developed to simulate both the aerosol generation by cutting fuel debris and aerosol removal by spray injection in the open-source CFD code OpenFOAM using the geometry of our model facility (UTARTS facility). Simulation results provide detailed information about the time evolution of aerosol distribution, and also show that aerosol generation rate has no influence on aerosol removal efficiency, and larger particles can be removed faster because of stronger inertial impaction. The newly developed numerical model with considering both aerosol generation due to cut fuel debris and aerosol scavenging by spray droplets can be used in the real-size containment vessel of Fukushima Daiichi nuclear reactors to optimize the design of future spray systems.
Sensitivity Calculation of Lower Shell Pressure Parameters in Advanced Nuclear Power Plant Accidents
Authors: Lanyu Zhou , Yanyu Sun , Yuntao Zheng , Changjiang Yang 
Published: November 2022

Abstract
As the main coolant piping from the steam generator to the reactor pressure vessel, cold leg is an important part of the primary circuit. During a LOCA accident in the cold leg, the parameter sensitivity has a crucial influence for the containment response. PAREO program is used to analyze the sensitivity of the pressure response of HPR1000 containment. Quantitatively evaluated the influence of free volume, area of thermal components, spray flow rate and internal atmospheric temperature on the pressure and temperature of containment under the condition of LOCA. The results show that the area of thermal components is the most sensitive parameter to the pressure response process of containment. The research results could provide technical support for containment design, safety analysis and assessment.
Information Security Design of Nuclear Power Embedded Real-Time Operating System
Authors: Li Meng , Shi Guilian , Ma Zhonggang , Wang Xiaowei , Fan Lichen , Dou Weiwei 
Published: November 2022

Abstract
With the development of modern automatic control technology, digital control technology has advanced technology and inherent advantages compared with traditional analog control technology. Therefore, the use of digital instrument and control system has become an inevitable trend of global nuclear power construction. Nuclear power embedded real-time operating system is an important part of digital instrument and control system in nuclear power plant. It is responsible for managing the hardware resources of the whole system and providing operation environment for application software. Its safety directly affects the safety of instrument and control system. This paper studies the application of embedded information technology in the safety of nuclear power plant, and puts forward a safety related scheme of embedded information system. The scheme fully considers the confidentiality, integrity and availability of information in the digital instrument and control system of nuclear power plant. The application results show that the scheme can meet the information security requirements of nuclear power instrument and control system, and has reference and guiding significance for the research and development of nuclear power embedded real-time operating system.
Classification Method and Response Strategy of Public Opinion Risk Related to Nuclear-Related Event Occurred in Countries Around China
Authors: Zhao Hongru , Tong Jiejuan 
Published: November 2022

Abstract
Due to the particularity of the nuclear-related event, even if it does not have any serious consequences, or even does not occur in China, it may also bring a greater risk of public opinion, that is, the technical risk is low but the public opinion risk is high. In order to provide a scientific basis to identify the impact of Nuclear-related Event occurred in countries Around China (NEAC) on China, we established an integrated classification framework of public opinion risks related to these events, and the indicators of each evaluation dimension have been identified. In another aspect, based on the theory of crisis life cycle and the theory of crisis management 4R, some suggestions on the response strategy of the public opinion risk caused by NEAC were proposed in this paper.
Microstructure and Hardness Investigation on 25Cr2Ni2MoV Steel Welding Joint of Nuclear Power Steam Turbine Low Pressure Welded Rotor and Engineering Applications
Authors: Huo Xin , Liu Xia 
Published: November 2022

Abstract
Large Nuclear power steam turbine Low pressure (LP) rotor composed of several shafts and discs with 25Cr2Ni2MoV steel was manufactured by gas tungsten-arc welding (GTAW) for backing welding and submerged-arc welding (SAW) for filling welding. In this study, the welding joint is investigated and it is found that the main microstructure of weld metal and base metal is typically bainite, while tempered martensite is observed in heat affected zone (HAZ) by over-quenching imposed by welding cycling and post weld heat treatment (PWHT). Compared with the weld metal of SAW, the microstructure of GTAW is relatively fine and homogeneous, due to its lower heat input. The hardness test indicates that the maximum hardness of HV400 occurs in HAZ because of tempered martensite. Weld metal of SAW has the hardness of HV270, which is a little bit lower than GTAW’s on account of fine grain. Based on the researches, the trial welded rotor and production rotor have been successfully produced, relevant inspections and tests were conducted, the results are well consistently with original research.
Neutronics Analysis of RBMK-1000 Reactor Using OpenMC: Criticality and Depletion Analysis at 3D Assembly Level
Authors: Hashemi Mohammad , Macián-Juan Rafael , Ohlerich Martin 
Published: November 2022

Abstract
This article presents the first results of the neutronics analysis of RBMK-1000 at the 3D assembly level using the OpenMC code as an early stage of the long-term large academic research project. The goal of the large academic research project is to develop a Multi-group Monte Carlo computational module around the OpenMC code using Python programming language. The results of this paper include the criticality calculation and depletion analysis of RBMK-1000. The effective multiplication factor value (keff) has been determined. In depletion calculation, the fission reaction rates of U-235, U-236, U-238, production of fissionable materials, and two fission product poisons Xe-135 and Sm-149 have been investigated. There is ongoing work for benchmarking the first results from OpenMC against Serpent 2 Monte Carlo code. The parametric design data of RBMK-1000 have been taken from the OECD - Nuclear Energy Agency (NEA) SFCOMPO 2.0 (Spent Fuel Isotopic Composition) database that includes measured isotopic concentrations of spent nuclear fuel, with operational histories and design data. The case study was the Leningrad-1 NPP. The ENDF/B-VII.1 data library has been used in this analysis. Moreover, this is the first time to date that the OpenMC code has been used in neutronic studies of the RBMK reactors.
Numerical Study on Welding Deformation of Upper Support Structure in 600MW Sodium-Cooled Fast Reactor
Authors: Dong Wenchao , Lu Shanping 
Published: November 2022

Abstract
The upper support structure of 600 MW sodium-cooled fast reactor is welded by inner and outer cylinders, upper and lower tube-sheets and multiple rib plates. The structure has a large size and many welding passes, which poses a great challenge to the control of welding deformation. It is an effective method to control welding deformation of structural parts by using numerical simulation method to predict welding deformation and optimize welding process. In this work, the welding process of typical T-joint is numerically simulated and tested firstly. The reliability of the simulated results is verified by comparing the temperature field, and the inherent strain of typical joint is obtained. Then, the inherent strain method is used to calculate the welding deformation of typical T-joint. The welding deformation accuracy is compared and analyzed by using thermoelastic-plastic method and inherent strain method, respectively. Finally, the welding process of the upper support structure is simulated by the inherent strain method. The influence of the designed welding sequence on the welding deformation of the upper support structure is studied, and the shrinkage allowance are obtained. The research results have been verified by field application, which provides technical support for the construction of 600 MW sodium-cooled fast reactor.
Design and Investigation of Liquid Lead-Bismuth Test Irradiation Loop in Research Reactor
Authors: Xu Yuheng , Si Juping , Zhang Liang , Sun Sheng , Wu Hongwei , Lu Mengkang 
Published: November 2022

Abstract
The Lead-bismuth eutectic (LBE) test irradiation loop is an indispensable testing vehicle for the investigation of LBE-cooled reactor fuel. To make a water-cooled research reactor have the irradiation test capacity of LBE-cooled reactor fuel, a test irradiation loop is investigated and preliminarily designed by absorbing the design and operation experience of the domestic and international heavy liquid metal loop. Structural material corrosion, impurity blockage, polonium diffusion, and metal solidification in the loop are considered in the design process. Corresponding measures have been taken to ensure the safe operation and emergency capacity of the loop. The loop is a 1500KW facility consisting of a coolant system and auxiliary systems the other, operating in a wide temperature range of 250–480 °C, oxygen concentration 0.01–0.05 ppm. The coolant system is to remove the heat from the fuel surface due to irradiation. Auxiliary systems are for coolant chemistry control, emergency fuel cooling, polonium control, and parameter measurement. In this paper, the technological process, main design parameters, and system composition of the loop are described, and the parameters and selections of main equipment, instrument parameters are proposed. The investigation supplies a useful reference for further design of test irradiation loop systems and equipment.
Nuclear Cognition and Public Acceptance
Authors: Zhu Weiqiang , Wang Peng , Fu Changying 
Published: November 2022

Abstract
Nuclear power generation is the main form of human use of nuclear energy in the period of peaceful development. Compared with the most widely used thermal power generation, nuclear power generation is safe, efficient and clean. Nuclear power generation is an inevitable choice for low-carbon development. However, the inherent radioactivity and leakage risk of nuclear power generation make the public full of doubts about its safety. This paper investigates the relevant content of nuclear power generation through the method of online questionnaire survey, and the survey results show that the risk of radioactivity and leakage in nuclear power generation is the main problem point of public concern, the more fully the public understands nuclear energy, the more scientific and accurate the understanding of the risks of nuclear power generation, and the greater the possibility of showing an optimistic attitude towards the development prospects of nuclear energy. Similarly, raising people’s nuclear awareness through publicity methods such as knowledge popularization, organizing visits, and online media can significantly increase the acceptance of nuclear power generation by the surrounding people. From the historical perspective of the development of human society, nuclear energy as a promising clean energy, on the basis of the continuous innovation of global nuclear technology, coupled with the continuous improvement of the public’s scientific literacy, the general public can better understand nuclear energy, accept the use of nuclear energy, nuclear power generation will also make greater contributions to global environmental protection and green development.
Research on Wastewater Source Term in High-Temperature Gas-Cooled Reactor and the Environmental Impact at Inland Sites
Authors: Zhang Yichun , Li Chuan , Li Hong , Cao Jianzhu 
Published: November 2022

Abstract
High-temperature gas-cooled reactor (HTGR) is paid more attention in terms of inherent safety, high efficiency, and less liquid radioactive waste. The research on the source term of wastewater and the discharge methods of liquid effluent from HTGR can not only reduce the possible radiation risk to the environment and the public, but also promote the construction of inland nuclear power plants (NPPs), and improve the public acceptance of inland nuclear facilities. In this paper, calculation of activated corrosion products (ACPs) in the cooling water systems have been done in HTR-PM (High-Temperature Gas-Cooled Reactor Pebble-Bed Module) under normal operation conditions. Besides, the discharge methods of liquid effluent from inland NPPs without receiving water or with poor dilution conditions are studied. The calculation results indicate the cooling water of the heat removal system and shielded and component cooling water system needs to be changed once during the 40-year lifespan. The cooling water of the reactor support cooling system does not need to be changed. The annual effective dose resulting from liquid effluent through liquid-to-gas discharge using the data from a typical inland NPP site in China is less than direct discharge into the receiving water. The liquid-to-gas discharge is suitable for inland NPPs without receiving water or with poor dilution conditions.