bluuebunny/crossref_metadata_2025_split huggingface dataset:
Development of a Synthetic Surrogate for F-24 From Blends of Iso-Paraffinic Kerosene (IPK) and Fischer-Tropsch Synthetic Kerosene (S8) in a Constant Volume Combustion Chamber (CVCC)
Authors: Soloiu Valentin , Parker Lily , Smith Richard , Weaver Amanda , Brant Austin , Rowell Aidan , Ilie Marcel 
Published: November 2022

Abstract
Investigations were conducted using mass blends of Iso-Paraffinic Kerosene (IPK) and Fischer-Tropsch Synthetic Kerosene (S8) to produce a synthetic surrogate for aerospace F-24. Due to the fossil fuel origin of F-24, the introduction of a synthetic surrogate would create a sustainable aviation fuel (SAF) with sources obtained from within the United States. An analysis of ignition delay (ID), combustion delay (CD), derived cetane number (DCN), negative temperature coefficient (NTC) region, Low-Temperature Heat Release region (LTHR) and High-Temperature Heat Release (HTHR) was conducted using a PAC CID 510 Constant Volume Combustion Chamber (CVCC). The fuels examined in this study are neat IPK, neat S8, neat F-24, and by mass percentages, as follows: 75IPK 25S8, 52IPK 48S8, 51IPK 49S8, 50IPK 50S8 and 25IPK 75S8.
The DCN values determined for IPK, S8, and F-24 were 26.92, 59.56 and 44.35 respectively. The influence of IPK present in the blends increases CD, thus reducing the DCN significantly. The fuel blend of 50IPK 50S8 was observed to be the closest match to F-24 when comparing DCN, ID and CD.
The surrogate blends were determined to have a lower magnitude of peak pressure ringing compared to that of the neat S8 and F-24, this is due to the extended NTC region caused by the IPK present in the blend. During further refinement of the surrogate blend, the Apparent Heat Release Rate (AHRR) curve for the 51IPK 49S8 fuel blend was found to have the closest match to the AHRR of F24. The surrogate blend 50IPK 50S8 was shown to have the smallest percent difference and best match during the LTHR stage, compared to F-24, while 52IPK 48S8 had the smallest percent difference for the energy released during LTHR. The ID and CD of the 25/75% blends were too dissimilar from the F-24 target to be considered as a surrogate.
A Noise Vibration Harshness (NVH) analysis was also conducted during the combustion of the three neat fuels in the CVCC. This analysis was conducted to relate the ID, CD, HTHR and ringing to the vibrations that occur during combustion. Neat S8 was observed to have the most vibrations occurring during the combustion process. Additionally, the HTHR was observed to have a distinct pattern for the three neat fuels and the combustion of these fuels was quieter overall.
Machine Learning and Genetic Algorithm Method for Powertrain Development: Rapid Generation of Engine Calibration Maps
Authors: Williams Zachary L. , Meruva Prathik , Bitsis Daniel Christopher 
Published: November 2022

Abstract
Meeting regulatory and customer demands requires detailed powertrain calibration which can be expensive and time-consuming. There is often a reliance on mathematical optimization tools to convert experimental learnings into a final calibration. This work focuses on developing multiple neural network machine learning (ML) models which were trained on different test-train data splits of test-cell recorded steady-state medium-duty (MD) diesel engine data. The output data was used to develop engine actuator maps by utilizing a genetic algorithm (GA). The genetic algorithm contains a fitness function which was varied to target different combinations of low NOx and CO2 emissions. The input variables used for the ML model were engine speed, engine torque, fuel rail pressure, exhaust gas recirculation (EGR) valve command, main injection timing, and wastegate valve command. The output variables predicted were NOx mass flow rate, exhaust temperature, fuel flow rate, and dry intake mass flow rate. The ML models were used to predict cycle-averaged engine-out emissions and time-series predictions of all output variables for different transient drive cycles. The drive cycles used for this case were the Heavy-Duty Federal Test Procedure (HDFTP) transient cycle, the Non-Road Transient Cycle (NRTC), the Ramped Mode Cycle (RMC) and the newly proposed on-road Low-Load Cycle (LLC).
In Situ Local Void Fraction Measurement in the Vertical Air-Gas Two-Phase Flow by the Combination of Multiple Current-Voltage (MCV) and Plural Long Short-Term Memory Lasso (MCV-pLSTM-L)
Authors: Saito Daisuke , Prayitno Yosephus Ardean Kurnianto , Sejati Prima Asmara , Takei Masahiro 
Published: November 2022

Abstract
In situ local void fraction in the vertical air-gas two-phase flow has been detected by the combination of multiple current-voltage (MCV) and plural long short term memory lasso (MCV-pLSTM-L) for the flow status monitoring in cooling equipment of nuclear reactor. The combination of MCV and machine learning has been applied to void fraction measurement of vertical air-gas two-phase flow. However, in our previous research Plural Long Short Term Memory Lasso, the machine learning employed drift flux model to calculate the true void fraction. There was a possibility of disagreement between true in situ void fraction and calculated void fraction. MCV-pLSTM-L has five steps, 1) true local void fraction measurement by electrical probe sensor, 2) voltage measurement by the MCV system, 3) voltage extraction by Lasso, 4) flow regime identification by 1st LSTM, and 5) void fraction α estimation by 2nd LSTM. Experiments were conducted in a vertical pipe with an inner diameter of 25mm and a length oh 80mm from the inlet. Two vertical flow regimes which were liquid single-phase flow and bubbly flow were used. As a result, the local void fraction is successfully estimated and applicability of MCV-pLSTM-L is confirmed.
Evolution of the Evening Method of Coolant Flow in the Large Core Based on Patent History
Authors: Yu Wenchi , Fu Hao 
Published: November 2022

Abstract
Generally, the reactor vessel is cylindrical having a hemispherical lower end. The core barrel is connected to the interior walls of the reactor vessel at or adjacent to the area where the cylindrical and hemispherical portions of the reactor vessel meet. Below the main core support, i.e. the core barrel that is capped at its lower end with the lower core support, the hemispherical vessel defines a lower head or lower plenum. A generally annular downcomer surrounds the reactor core barrel between the core barrel and an inner wall of the reactor vessel. Cooling fluid, typically water, is pumped into this annular downcomer. The coolant fluid circulates downward into the lower plenum. The hemispherical shape of the lower plenum assists in evenly circulating the coolant fluid therein. A plurality of reactor core coolant inlet openings are located on the underside of the lower core support plate. Coolant flows from the lower plenum, into the core coolant inlet openings and upwardly into the core to cool the fuel assemblies.
With the advent of larger plants with larger cores it became evident that further means were necessary to improve the distribution of coolant flow in the lower plenum to assure uniform coolant flow and pressure were maintained across all of the reactor core coolant inlet openings in the lower core support plate.
Non-uniform coolant pressure or flow causes uneven coolant flow into the core, which results in uneven cooling of the fuel assemblies of the core. Uneven fuel assembly cooling may force the entire core to be derated to accommodate “hot assembly” locations. Non-uniform coolant flow and pressure may result in vortices or other flow disruptions to form in the coolant fluid circulating in the lower plenum. According to patent filing history, dozens of devices have been proposed to uniform the distribution of coolant flow. Those device could be categorized several kinds. As coolant flow in the fringe of low core plate, firstly the cylinder shape has been applied to prevent the vortex. This biggest vortex is formed by that the downflow from core barrel has 180° direction diverting with the upflow into LCP, diverting between the high speed flows. In order to maintain adequate and uniform cooling throughout the core, it is important that a uniform coolant flow and pressure be maintained across all of the reactor core coolant inlet openings in the lower core support plate. The second way to is put some device in the bottom of lower plenum to prevent vortex. The third is changing flow distribution of LCP by setting some flow limiting devices.
Application of First-Order Method to Estimate Structural Integrity in a Probabilistic Form of Component Subjected to Thermal Transient for Optimization of Design Parameter
Authors: Okajima Satoshi , Mori Takero , Kikuchi Norihiro , Tanaka Masaaki , Miyazaki Masashi 
Published: November 2022

Abstract
Japan Atomic Energy Agency has been developing “Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA)”, which aims to offer best solution for challenges on the design and operation of a nuclear plant. In a part of ARKADIA for design study including design optimization of components named ARKADIA-Design, we have been developing a process to automatically optimize design parameter of structural components subjected to various kinds of loads, including thermal transient.
In this paper, we propose the simplified procedure to estimate failure probability of components subjected to thermal transient for the design optimization. An objective function of this optimization based on failure probability of the components is assumed, because failure probability can be commonly used as an indicator of component integrity for various failure mechanisms. In order to enable necessary number of the estimations for the design optimization with practical calculation cost, we have aimed to reduce number of analyses required for one estimation. For this purpose, we have adopted the First Order Second Moment (FOSM) method as the estimation method of failure probability on the process of the optimization and an orthogonal table in experiment design method is utilized to define conditions of the analyses for the evaluation of mean and variance of thermal transient stress, which are used as inputs for the FOSM method. The proposed procedure has been applied to the demonstration study to optimize thickness of cylindrical vessel subjected to thermal transient derived from shutdown. We confirmed that the procedure can evaluate the failure probability depending on the cylinder thickness with practical calculation cost.
Research on Generalization of Typical Data-Driven Fault Diagnosis Methods for Nuclear Power Plants
Authors: Li Jiangkuan , Lin Meng 
Published: November 2022

Abstract
With the development of artificial intelligence technology, various achievements have been realized in data-driven nuclear power plant fault diagnosis. Even endowed with high flexibility and practicability, most of the proposed data-driven methods are based on the same assumptions that the test data is in the same distribution as the training data. In practice, nuclear power plants may be in variable operating conditions, which brings challenges to the generalization of the diagnosis model trained by finite data. In this paper, the widely used data-driven models in nuclear power plant fault diagnosis: Random Forest (RF), K-Nearest Neighbor algorithm (KNN), Fully Connected Neural Network (FCNN) and Convolutional Neural Network (CNN) are taken as examples to study the influence of the distribution discrepancy between training data (source domain) and test data (target domain) on their generalization. The results show that the distribution discrepancy exert serious adverse effects on the diagnostic performance of the data-driven models. At the same time, to improve the generalization of data-driven models, a nuclear power plant fault diagnosis transfer learning method based on pre-trained model is proposed, which can utilize the fault diagnosis knowledge from the source domain task to accelerate the model training in the target domain task, so that the model can achieve better diagnosis performance with limited labeled data in target domain.
Creep Performance and Microstructure Evolution of IN617 at 950 °C for VHTR Applications
Authors: Wang Yue , Wang Haitao , Li Kejian 
Published: November 2022

Abstract
IN617 was considered the primary material candidate for the IHX in the VHTR. Researching microstructure evolution during high-temperature creep of IN617 helped understand its fracture laws and guide IHX operation under creep loading. Creep tests of IN617 were conducted under 19 MPa, 24 MPa, 27 MPa, and 38 MPa at 950 °C. Creep rupture mechanisms of IN617 were discussed by correlating creep performance, microstructure characteristics and fracture-surface morphology. The results indicated that DRX, creep voids and brittle-phase precipitation were found under different stresses during microstructure observation, which would cause the specimen ductile rupture, intergranular rupture and brittle rupture, respectively. Specifically, under the highest stress 38 MPa, DRX occurred and grain size was decreased greatly to 11.1 μm from 99.7 μm. Fine grains were easy to migrate, causing significant plastic deformation and ductile rupture of specimens. With stress decreased to 27 MPa, grain boundaries became vulnerable and intergranular rupture occurred because intergranular carbides dissolved and their pinning effect was weakened. As stresses were lowered to 24 MPa and 19 MPa, nitrogen was diffused into specimens and brittle nitrides precipitated into continuous networks along GBs. The internal cracking of nitride networks caused brittle rupture. Meanwhile, steady creep rates were increased, and creep rupture lives were shortened greatly, especially under 19 MPa.
Preliminary Study on Corrosion Behavior and Mechanical Properties Of Inconel 617 in Impure Helium Environment of VHTR
Authors: Du Bin , Yin Huaqiang , Ma Tao , Li Haoxiang , Zheng Wei , He Xuedong 
Published: November 2022

Abstract
Helium is generally used as coolant in Very-High-Temperature Reactor (VHTR), but a small amount of impurity gas will inevitably be mixed in the primary coolant during construction, operation and maintenance. At high temperature, these impurity gases will corrode with the alloy materials of the intermediate heat exchanger, resulting in the decline of the properties of the superalloy materials. This paper mainly studied the changes of microstructure and mechanical properties of Inconel 617, a candidate material for high temperature reactor intermediate heat exchanger, after aging for 50 hours in impure helium at 950 °C. The microstructure of the alloy was characterized by weighing, scanning electron microscope and energy dispersive X-ray spectroscopy. After corrosion, Cr oxide layer was formed on the alloy surface, and Al internal oxidation appeared below the oxide layer. The mechanical test results showed that the strength and plasticity of Inconel 617 decreased significantly, which was related to the intergranular oxides and carbides. The fracture morphology of the alloy is mainly brittle fracture.
A Bi-Layered Three-Dimensional Mechanical Modeling of the Cladding and Its Creep Deformation Analysis
Authors: Zhang Ming , Li Jinggang , Liu Xiaohan , Lu Yong , Zhu Yanan 
Published: November 2022

Abstract
The cladding of the fuel rod is a cylinder-shaped structure made of Zirconium alloy, which will collapse due to structural creep under the extremely service conditions such as high temperature, high pressure, and high irradiation. The collapse of the cladding results in losing its structure function and threatening the safety of the reactor. Based on the commercial finite element software ABAQUS with its user subroutine CREEP, a bi-layered (coating and matrix material) three-dimensional (3D) cylindrical cladding model is established with thermal and irradiation finite creep behavior. The external pressure is assumed to be constant acting on the outer surface. The deformation and the rate of deformation increase with the increasing of the irradiation time in the reactor, which leads to the collapse of the cladding eventually. The initial ovality has a positive effect on the creep deformation. Compared with the single-layered model, the coating of the bi-layered cladding can prevent the Zirconium alloy matrix from excessive creep deformation and thus can protect the cladding. The thicker the coating, the stronger the protective effect from the mechanical point of view. A qualitative case of the cladding creep burst was simulated, and the behavior of the creep burst and creep collapse is similar. The corrosion and oxidation behavior are not considered herein for simplicity. The current bi-layered 3d model can be extended to the structural design, safety analysis, as well as life evaluation of some multi-layered cladding of the accident tolerant fuel (ATF).
Research on Risk Assessment of Heavy Load Drops in Nuclear Power Plants
Authors: Ma Chao , Wang Yiming , Feng Churan , Yang Jian 
Published: November 2022

Abstract
After years of development, probabilistic safety assessment has become a tool for the evaluation and optimization of the design and operation of nuclear power plants, and the analysis scope has also developed from internal events to internal hazards and external hazards. In the assessment of internal and external hazards, the probabilistic safety assessment of internal fire, internal flooding and seismic has been fully studied, and a systematic analysis method has been formed, which is used to support the design and operation of nuclear power plants. The drop of heavy load is one of the internal hazards of nuclear power plants. At present, there is no systematic method how to quantitatively evaluate the risks to nuclear power plants. This paper refers to the analysis requirements of IAEA, NEI and other institutions on the analysis of heavy load drop, and combines the design characteristics and operating experience of the power plant to carry out a study on the risk assessment of heavy load drop, including: analysis of the possibility of heavy load drop, analysis of the impact of heavy load drop, CDF levels caused by heavy load drop, etc. Through the research in this paper, it is expected to provide insights into the risk of heavy load drop and provide support for subsequent risk assessment and optimization of nuclear power plants.
Probabilistic Safety Assessment of Newly Added Auxiliary Transformers In Fangjiashan Nuclear Power Plant
Authors: Zou Shengjia , Wu Jian , Cao Yong , Wang Zilong , Luo Yang , Zhao Ming , Liu Deyi , Chen Honghao 
Published: November 2022

Abstract
In order to further improve the safety of the nuclear power plant, China nuclear power Fangjiashan nuclear power plant plans to add an auxiliary standby transformer, which is used for the redundancy of the original auxiliary transformer. This permanent change will improve the reliability of external power supply of nuclear power plant and reduce the frequency of loss of external power supply, so as to improve the safety level of Fangjiashan nuclear power plant. In this paper, the probabilistic safety assessment method is applied to analyze Fangjiashan unit, and the level-1 PSA model of internal events under power operation conditions, level-1 PSA model of internal events under low power and shutdown conditions and spent fuel pool model are modeled respectively. After the permanent change of auxiliary standby transformer is added during Fangjiashan power operation, low power and shutdown, The quantitative safety analysis results are given for the improvement of safety and reliability of Fangjiashan nuclear power plant. This evaluation provides an idea for the application of PSA in the permanent change of nuclear power plant. Before the permanent change of power plant plan, through the analysis of probabilistic safety evaluation, we can get how much benefit this change brings to the nuclear safety of power plant.
Research on Deposition Performance of Fine Particles of Heat Pipe in Small Reactor
Authors: Liu Xiaofang , Zhou Tao , Mao Shang , Wei Dong 
Published: November 2022

Abstract
Small modular reactor (SMR) adopts heat pipe as the heat exchanger, which can improve its inherent safety. In the process of heat exchange, the deposition of fine particles caused by fluid flowing through the surface of heat pipe will cause heat transfer deterioration and the corrosion of pipe wall, which will adversely affect the safe operation of the reactor. The law of fine particle deposition on the outer wall of the simplified heat pipe model was numerically simulated, and the particle deposition in the fluid field outside the pipe was obtained under different temperature difference and flow velocity. The results showed that in the process of axial flow along the pipe wall, the velocity near the wall decreases gradually along the radial direction, and decreases to a fairly small value at the pipe wall; due to the heating effect of the heat pipe, the temperature of the fluid near the wall increases gradually along the axis, and there is a heat affected zone in the fluid temperature on the cross section. With the increase of working fluid flow rate and inlet temperature, the particle deposition rate increased first and then decreased.
On Line Monitoring of Sensor Failure in Nuclear Power Plant Based on Sequential Probability Ratio Test
Authors: Yu Fangxiaozhi , Ma Xintong , Miao Zhuang , Zhang Min 
Published: November 2022

Abstract
The safe and reliable operation of nuclear power plant is inseparable from the integrity of the control system function, and the integrity of the sensor function is the basis for the I&C system to monitor the operation status of nuclear power plant and trigger automatic action. However, there are some shortcomings in the traditional threshold alarm method of nuclear power plant: for example, it is difficult to identify the tiny sensor failure (such as tiny drift) earlier by this method. The threshold of this method is fixed and cannot be adjusted according to the actual situation (such as transient change). This paper introduces the on-line monitoring technology of sensor failure, which plays an important role in reducing the probability of unplanned shutdown and unit design and operation cost. In the future, the on-line monitoring technology of sensor failure has a good application prospect, and the relevant research is of great significance.
This paper mainly studies an overall technical scheme of sensor failure monitoring in nuclear power plant. In this paper, the data of the PWR nuclear power plant simulation simulator are used, and the sequential probability ratio test (SPRT) model, a typical failure diagnosis algorithm, is used to establish the sensor condition monitoring model, and the performance of the monitoring model are tested. The results show that the SPRT sensor monitoring model established in this paper has high accuracy, and can effectively monitor the occurrence of tiny drift phenomenon of sensors, and preliminarily verify the feasibility of on-line monitoring technology of sensor failure in PWR nuclear power plant.