bluuebunny/crossref_metadata_2025_split huggingface dataset:
Spreading Behavior of Molten Metal on Flat Plate in a Shallow Water Pool
Authors: Yamamoto Yasunori , Ito Tomomasa , Nihashi Kyosuke , Miwa Shuichiro 
Published: October 2021

Abstract
The spreading behavior of the molten core is an essential phenomenon to consider for understanding the initial conditions of molten core-concreate interaction and the coolability of the core. In this study, experiments using molten tin were performed to obtain the necessary data to analyze the spreading behavior of the molten core under wet conditions. The experimental parameters used were the nozzle diameter, water depth, and molten metal temperature. The pressure waves occurred due to the molten metal-water interaction in the water pool, especially under high molten metal temperature conditions. They were observed during the floor collision process and the spreading process after the collision. Our results indicated that the spreading area and the shape of the solidified metals strongly depended on the pressure waves due to the thermal interaction between the molten tin and water. When the interface temperature between the molten tin and water was lower than the homogeneous nucleation temperature, the pressure wave rarely occurred. The effect of the pressure wave on the spreading areas increased with the superheat of the molten tin.
Theoretical Study on Single Bubble Expansion After Pressurized Subcooled Water Injected in Thermal Glycerin
Authors: Mao Feng , Zhang Lei , Yun Xiangyu , Lu Donghua , Tian Wenxi , Zhang Huiyong 
Published: October 2021

Abstract
In the hypothetical liquid metal reactor SGTR (Steam Generator Tube Rupture) accident, the high-pressure water in steam generator tube will be injected into the high-temperature liquid metal. During the water-liquid metal interaction, water vaporized and vapor bubbles migrated into liquid metal, which could be resulting a sharp pressure raise and threatening the safety of the reactor. This multi-phase system can be abstracted as a system in which a high-pressure subcooled fluid (overheating state at atmospheric pressure) is injected into another higher temperature fluid (normal pressure) to form a two-component (water and liquid metal) system in which multiphase (fluid and vapor) were existed.
There are two effect could cause bubbles expansion which lead to overall system pressure increase, they are flash evaporation of pressurized water since the depressurization and high-temperature glycerin heating the water/vapor. In this paper, the theoretical analysis of the single bubble expansion model is used to evaluate the effect of these two phenomena on the system pressure rise. First of all, this paper uses the force balance point of view to derive the single bubble expansion equation inside the inviscid fluid, and obtain the relationship between bubble expansion speed and bubble internal pressure. Secondly, through conservative assumptions, the bubble expansion speed caused by the heating effect of the high temperature fluid on the bubble is evaluated. Finally, this paper compares the strength of these two effects in the expansion of a single bubble, and theoretically explains the driving force of the pressure increase in the simulation experiment and the potential flow pattern.
Study on Aging Management of Operating Nuclear Power Plants in China
Authors: Liang Li , Guixiang Yi 
Published: October 2021

Abstract
In 2016, The first nuclear power plant submitted an application for renewal of operation license to National Nuclear Safety Administration (NNSA) in China. In order to guide the project, NNSA issued the technical policy for extension of the validity period of the nuclear power plant operation license, which mainly refers to the US nuclear power plant license renewal review policy. However, China’s nuclear safety regulation system mainly refers to the establishment of the International Atomic Energy Agency (IAEA), and carries out 10-year Periodic Safety Review for nuclear power plants in the actual operation process. There are some differences between the U.S. license renewal strategy and the IAEA long-term operation system. An important part of both is aging management. Based on the daily operation and supervision of China’s nuclear power plants, this paper summarizes the differences and similarities of aging management between periodic safety review and operation license renewal application, and puts forward the contents that should be added and improved in the periodic safety review, so as to effectively link up the application system for operation license renewal and the Periodic Safety Review system, to revise the “Periodic safety review of nuclear power plant” (HAD103/11) and “Aging management of nuclear power plant” (HAD 103/12) provide reference.
Preliminary Design of a Fuel Element With Divergent Hot Gas Channel In Particle Bed Reactor for Nuclear Thermal Propulsion
Authors: Liang Zhaoyu , Ji Yu , Sun Jun , Mao Chenrui , Shi Lei 
Published: October 2021

Abstract
The nuclear thermal propulsion (NTP) system can shorten the travel time in deep space exploration and reduce the initial mass of the launch vehicle due to its superior characteristics including high specific impulse and large thrust. Particle bed reactor (PBR) is one of the most appropriate reactor concepts to equip the NTP systems. To make the best use of PBR, the thermal-hydraulic design of the fuel element should be carefully considered and a flow-power matching technology should be developed. In this paper, a novel design employing a divergent hot gas channel is proposed to achieve a uniform flow distribution with lower maximum temperature and pressure drop. Through the analysis of the 1D modified momentum equation in the inlet plenum and hot gas channel, the model of pressure drop is established. Then, the differential equation of the ideal cross-section of the hot gas channel is derived. At last, the flow and heat transfer process in the fuel element with divergent hot gas channel is simulated by using computational fluid dynamics (CFD) code, and the reduction of pressure drop and temperature verifies the theoretical model.
This study shows that the proposed design of the divergent hot gas channel can provide a new idea for thermal-hydraulic optimization of the PBR fuel element.
Preliminary Research on The Thermal-Mechanical Coupling Behavior Simulation Method of M3 Fuel
Authors: Changbing Tang , Yongjun Jiao , Yuanming Li , Kun Zhang 
Published: October 2021

Abstract
M3 fuel which microencapsulated TRISO particles in zirconium metal matrix is an integrated fuel designed to replace the traditional pellet-cladding fuel rod. The fuel design can effectively solve the problems of LWR fuel, such as water side corrosion, abrasion, pellet-cladding interaction, stress corrosion cracking and internal pressure bracing. Because of the multiple coating structure of TRISO particles and complex contacting behavior and deformation between TRISO particle and metal matrix, M3 fuel exhibits complex in-pile behavior. In order to establish the simulation method for the M3 fuel, the three-dimensional model was established by using ABAQUS software. The complex contacting behavior and deformation of coated layers and matrix for M3 fuel was considered. The results indicated that bonded strength between TRISO particle and Zr matrix has great effect on the stress distribution of SiC layer and matrix. The vulnerable area of Zr matrix located in the nearest position of TRISO particle to TRISO particle. The maximum tensile hoop stress of SiC layer was up to 150 MPa under strong bonding state, while SiC layer suffered compressive stress during all operation time for the weak bonding state. The failure probability of SiC layer was about 7 × 10−3 under strong bonding state which much higher than the weak ones. Thus, the integrity of SiC layer can be maintained by controlling the bonding strength between TRISO particle and Zr matrix during manufacturing process.
A New Insight Into Molten Corium Concrete Interaction With Concrete Ablation Analysis for Mitigation Scheme
Authors: Khurshid Ilyas , Alade Amidu Muritala , Addad Yacine , Afgan Imran 
Published: October 2021

Abstract
The study presents the use of CORQUENCH simulator to simultaneously model the molten corium, composition of concrete, molten corium heat transfer models and solve the related chemical reactions. Using this modeling technique, the chemical reaction capabilities of CORQUENCH were successfully utilized which enabled the modeling of interaction between molten corium and concrete. The developed model was validated against experimental data and the results showed that the temperature of corium, composition of concrete and water injection time have a pronounced effect on mitigating depth of ablation and reactor integrity in case of a nuclear accident. In addition, the composition of concrete is the main controlling factor to mitigate ablation in the investigated case study. An alternative to concrete, could be a certain igneous rock (tested in this study), can lead to comparatively low rates of ablation due to its high thermal resistant properties. Furthermore, the injection of water (as a cooling agent) into the reactor cavity should also be optimized to enhance corium quenching to avoid ablation via basement melt-through. As a consequence of concrete ablation by molten corium, the depth of ablation is expected to decrease to zero ablation with the suggested material in the selected case study. The concrete ablation mechanisms during MCCI are very case-dependent on the concrete solidus, liquidus and ablation temperature, respectively.
Exploring Probability of Gas Entrainment With CFD Analysis of the Flow in the MICAS Experimental Facility
Authors: Bhatia Harshit , Bieder Ulrich , Gorsse Yannick , Guenadou David 
Published: October 2021

Abstract
Investigation of sodium cooled fast reactors (SFR) is ongoing in France. There have been a number of experiments carried out within the framework of Generation IV reactor program at CEA. MICAS experimental facility is one of these experiments, which represents the hot pool facility of a SFR in a 1/6th scaled down model. Recent studies carried out in MICAS, explore the probability of vortex formation at one of the internal heat exchangers (IHX) nozzle, which can further lead to gasentrainment from the pool surface. Therefore, MICAS facility has extensively tested different flow conditions and the non-symmetric IHX operation.
Multi-scale calculations are carried out using Computational Fluid Dynamics (CFD) in order to better understand the vortex formation at the IHX plenum and the associated possible gas entrainment into the vortex from the hot pool free surface. This paper presents the computational analysis with focus on operation of the SFR, which tries to replicate the possibility of vortex formation, under non-symmetric operation of two out of four IHXs. The analysis was carried out with the open source code TrioCFD. This study provides valuable insight into possibilities, constrains and limitation of the used CFD calculation methodology.
The Evaluation of Break Sizes of LOCA by the Temperature Difference at the Recirculation Inlets of BWR
Authors: Hwang Sheng-Dih 
Published: October 2021

Abstract
In this study, the MAAP5 code was used as a tool to simulate a BWR6/MarkIII nuclear reactor and which was LOCA with different break sizes at the recirculation loops. By detecting the temperature difference between the two loops, and then the break size could be determined. The results were consistent with the previous experiments or simulations. The study could be suggested to the operators at the main control room (MCR), when they observed some syndromes of the data and they could preliminary estimate the size of the LOCA. Those could help the operators to determine a strategy to avoid the probability of severe accident.
Development of an Integrated Design Evaluation ‘HITEP’ Platform for High-Temperature Pressure Boundary Components and Piping Systems
Authors: Lee Hyeong-Yeon , Jeong Si-Hwa , Won Min-Gu , Huh Nam-Su 
Published: October 2021

Abstract
An integrated software platform of elevated temperature design evaluation for pressure boundary components and piping subjected to high-temperature operation in the creep range has been developed. The software of ‘HITEP’ platform is capable of performing high-reliability design evaluations according to two nuclear grade design rules of ASME BPVC Section III Subsection HB and RCC-MRx, and one non-nuclear grade the Section VIII Division 2 Code Case 2843-2. The program modules were verified with a number of example problems. The HITTEP platform can be used for a reliable and efficient elevated temperature design of pressure boundary components and piping systems in generation IV nuclear systems, supercritical thermal power plants and oil refinery plants subjected to high-temperature operation in the creep range.
Heat Transfer Analysis of a Conceptual Horizontally-Oriented High Temperature Gas-Cooled Reactor
Authors: Feng Jinyong , Lopez Enrique Velez , Stewart William Robb , Wiser Ralph , Baglietto Emilio , Shirvan Koroush 
Published: October 2021

Abstract
High temperature gas-cooled reactors (HTGRs) have been constructed around the world since the 1960s. Compared to light water reactors, HTGRs feature a low core power density, which ultimately results in the need for larger components and structures. The successful widespread deployment of HTGRs will likely require a significant reduction in reactor building size per kW, to compete with other sources of carbon-free energy. The Modular Integrated Gas-cooled High Temperature Reactor (MIGHT-R) has been recently proposed to leverage inherent safety characteristics of HTGRs and its proven technology, while increasing its reactor building power density by 4 to 5 times. Contrarily to the vertical side-by-side orientation of the reactor core and steam generator of the typical HTGR design, the MIGHTR primary system components are laid horizontally, inline with one another. In this paper, a design embodiment of MIGHT-R concept is presented and preliminary heat transfer analysis under normal operating conditions is performed adopting a simplified porous representation of the flow behavior inside the reactor core and steam generator. The predicted mass flow rates in the fuel region, inner reflectors and outer reflectors agree with analytical estimations. A set of sensitivity tests for components variations are presented to provide insights on the influence of design parameters. A preliminary design reactor cavity cooling system (RCCS) is presented which is basically a 100 °C “water shell” surrounding the reactor vessel and it only covers the reactor core regions. Comparison of the same design with different orientations is also made and indicates that the mass flow and temperature distribution inside the reactor core have low dependence on orientation. Based on the computational results, the stratification of the coolant in the horizontal configuration for normal operation is deemed negligible. The presented work supports the normal operation feasibility of a horizontally-oriented HTGR concept with its potential to significantly reduce HTGR capital cost.
Failure Strain Criterion Accounting for the Effect of Material Strength
Authors: Baghous N. , Barsoum I. 
Published: October 2021

Abstract
The objective of this study is to investigate the effect of the Lode parameter on different material strengths. Recent work has shown that ductile failure highly depends on the stress state characterized by both the stress triaxiality T and the Lode parameter L, which is related to the third deviatoric stress invariant. Thus, for six different steel grades, two different specimen geometries were manufactured to account for two different Lode parameters (L = −1 and L = 0), whereas T is controlled by introducing different sized notches at the center of the specimens. By performing tensile experiments and running finite element simulations, the ductile failure loci of the six materials showed variations between the two specimen geometries, indicating that the failure highly depends on the stress state characterized by both T and L. This indicates the need to reassess the ductile local failure criterion in the ASME codes that only accounts for T as a stress state measure. A Lode sensitivity parameter LS is defined based on the experimental results and revealed that the steel grades with ultimate strength higher than a certain threshold value (450 MPa) exhibit sensitivity to the Lode parameter, and the results showed that the LS increases with increase in the ultimate strength of the steel grade. The results were incorporated to enhance the original ASME local failure criterion by accounting for T, L, and LS to accurately assess ductile failure in high-strength steels.
Application of Ultrasonic Guided Wave in LMPH Tube of Ethylene Cracking Furnace
Authors: Ding Ju , Wang Shenghui , Liu Shuhong 
Published: October 2021

Abstract
In view of the frequent leakage and perforation of the steam system pipe fittings of a Chinese factory in a short period of use, it will not only affect the normal use of the pipe system, but also cause safety hazards. In this paper, sampling and analysis of failed pipe fittings are carried out. Through the analysis of macro and micro corrosion morphology on site, it is shown that local corrosion is the root cause of failure, and it is accompanied by uniform corrosion. When the corrosion reaches a certain level, such as perforation, the overall pipeline will increase larger system security risks. So, what can we do to prevent these accidents? In this article, we make some useful suggestions: Strict control of the quality of cooling water. Do a good job of steam defogging. Timely replace the pipe fittings with serious thinning and corrosion thinning. Change the material to ultra-low carbon stainless steel (316L) or Wear-resistant materials. Comprehensive inspection and structural integrity safety assessment to ensure the safe and effective operation of equipment.
Technical Basis for Proposed ASME Code Case for Construction of Compact Heat Exchangers in High Temperature Reactors
Authors: Keating Robert , McKillop Suzanne 
Published: October 2021

Abstract
The U.S. Department of Energy has recently completed a research program to support the development of the compact heat exchanger (CHX) for use in high temperature advanced reactors. The project was executed by an Integrated Research Project (IRP) and includes team members from the University of Wisconsin–Madison, University of Michigan, Georgia Institute of Technology, University of Idaho, North Carolina State University, Oregon State University, Electric Power Research Institute, MPR Associates, and heat exchanger manufacturers CompRex and Vacuum Process Engineering.
The research was guided by a roadmap to support the eventual adoption of an American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Code Case for compact heat exchanger designs in high temperature reactor service. This roadmap is outlined in Journal of Nuclear Engineering and Radiation Science, NERS 19-1161, which identified specific research gaps to support the Code Case. The IRP has completed the research and the ASME Section III Standards Committee is in the process of writing and approving a Code Case for Section III, Division 5, Class A applications.
This paper provides the detailed technical basis for the ASME Code Case for CHX. The paper provides a vital link between the research results and the proposed Code Case and closes the research gaps that are tied to the original Roadmap. In addition, the paper provides further detailed commentary on the proposed requirements. The major technology gaps included material properties, failure modes and effects, analysis methods, and examination methods. The IRP studied basic material properties of diffusion bonded plate, creep and fatigue models, development of NDE methodology and development of advanced analytical approaches to design.
Prediction Method and Application Case Analysis of Dangerous Parts of Pipelines in Petrochemical Plants
Authors: Fang Zhou , Chen Ruxin , Wang Zhe , Huan Zhibo , Liu Qia , Yan He 
Published: October 2021

Abstract
A prediction method of dangerous points resulting from dynamic erosion was studied by combining Computational Fluid Dynamics and application cases. CFD numerical simulation software was used to simulate conditions of real process in industry. Considering the factors like geometry, flow velocity, flow direction, different erosion results in various situation were exampled and analyzed. Working conditions were sorted out and related sample graphs were made and indexed. This prediction method can reveal flow law of internal flow field and flow field distribution. Hence, critical points caused by dynamic erosion can be found and sensor-layout plan for online inspection, monitoring and management can be introduced. The safety in the use of industrial pipelines can be guaranteed. This method had been applied on pipelines of an oil and gas company in western China. A series of studies regarding their erosion rates and trend predictions were done based on actual situation on site. The technical route for applying this method was explained in detail. The corrosion mechanism and calculation method for this case were concretely stated. The prediction results and influencing factors of erosion parts were stated and the significance of using above-mentioned prediction method is declared.
Application of Different Data-Driven Methods in Material Performance Prediction of 2.25Cr-1Mo-0.25V Steel After Forming and Tempering
Authors: Li You , Chen Zhiping , Li Hongfei 
Published: October 2021

Abstract
2.25Cr-1Mo-0.25V steel is the main material of hydrogenation reactor due to its good high temperature properties and hydrogen embrittlement resistance. As we all know, there is a great performance gap between virgin material and actual product. Nevertheless, current design methods cannot evaluate the impact of manufacturing. In engineering practice, tensile testing is widely implemented on online surveillance samples or experimentally simulated samples to acquire the actual material characteristics. However, sampling from service hydrogenation reactor, as well as simulating actual manufacturing process, is time-consuming, labor-intensive, and difficult. To systematically analyze the effect of fabrication on 2.25Cr-1Mo-0.25V steel, a prediction method that can consider manufacturing residual effects needs to be proposed. Nowadays, many data-driven methods, such as back propagation neural network (BPNN), support vector machine (SVM) and random forest (RF), are employed in predicting material behavior during hot forming, whereas they are scarcely applied on 2.25Cr-1Mo-0.25V steel. In this work, the mentioned three machine learning procedures were constructed to simulate the behavior of actually manufactured 2.25Cr-1Mo-0.25V steel in uniaxial tensile test. Firstly, tensile samples were extracted from different locations of an actually manufactured hydrogenation reactor, and stretched at 454 °C and 482 °C. Subsequently, BPNN, SVM and RF models were developed on the obtained tensile curves, respectively. Their inputs were manufacturing parameter, sampling location, service temperature and tensile strain, whereas their output was tensile stress. Finally, the predictability of models was evaluated by test set data which were not involved in model training process. The results show that the predictions of BPNN, SVM and RF all agree well with experimental data. BPNN exhibits lowest prediction error and highest generalization power, SVM is somewhere in between, while RF is the worst. This work will contribute to the application of data-driven methods in material property prediction of 2.25Cr-1Mo-0.25V steel.