bluuebunny/crossref_metadata_2025_split huggingface dataset:
Research on Nuclear Safety Video Display Unit Technology Based on Digital Twin
Authors: Wu Yanqun , Zhao Yang , Liu Mingxing , Jiang Wei , Chen Qi , Wang Shun , Yan Hao 
Published: October 2020

Abstract
The nuclear safety video display unit (SVDU) in the nuclear safety level DCS implements functions such as displaying and warning of reactor safety related parameters and sending safety control instructions, which belongs to safety level 1 equipment. As a high-cost complex safety level display device, due to its high functional complexity, high security and reliability requirements, and special usage scenarios, its design, research, and function verification have always encountered problems such as low intelligence and low digitization, resulting in slow development process, complicated product function verification, and inconvenient use and training, etc. After the SVDU is put into practical use, continuous analysis of its stability, reliability, and safety, and its health status monitoring and maintenance are also difficult technical problems.
Based on the five-dimensional digital twin model as a design benchmark, a digital twin-based design method for SVDU is proposed. Taking the SVDU in a nuclear safety level DCS (NASPIC) as the object, the digital twin technology is adopted to model the physical objects such as display unit, human-machine interface unit, storage unit and communication network unit, and the high-speed industrial Ethernet network is used to map and interconnect the data between the components, so as to realize the physical fusion, model fusion and data fusion of the real SVDU and the virtual SVDU. With the help of data feedback from safety level DCS, the data, symbol display and control process of SVDU can be visualized and analyzed in virtual environment, and the real-time control function verification, fault early warning and auxiliary decision-making can be carried out, which improves the visibility, accessibility, operability and predictability of real SVDU display and control process. The real-time data, historical data, fault self-diagnosis data, and expert experience of real SVDU and virtual SVDU are incorporated into the twin data pool to reproduce the variable-speed replay of the historical operation process of SVDU and realize the post-accident condition analysis with multiple spatiotemporal dimensions; the record and analysis of self-fault diagnosis data provides the possibility of stability analysis services such as clock, power supply, human-machine input, and storage equipment, etc.
Digital twin-based SVDU technology can ensure rapid development and iteration of products under the requirements of complex functions and high safety and reliability, and truly reproduce the display and control effects of SVDU, while meeting users’ multiple application scenarios and data service requirements.
Study on the Thermal-Mechanical Performance of SiC Composites Cladding Under Multiple Conditions
Authors: Yin Chunyu 
Published: October 2020

Abstract
SiC has become a candidate cladding material of Accident Tolerant Fuels (ATF) due to its excellent irradiation stability and corrosion resistance. However, because SiC is a ceramic material with low toughness, brittle failure is a significant concern. In order to improve the toughness, SiC fiber is required to manufacture multi-layer SiC composites. But the current performance model or analysis tool is not available for SiC composites cladding due to its obviously difference with Zr alloy cladding. On one side, Finite element method was used in this paper to analyze the performance of SiC composites cladding under operation conditions which include normal, transient conditions and LOCA conditions; on the other side, this paper gives the performance of the SiC composites with two layers under multiple operating conditions. The result showed that the temperature was stable and the maximum hoop stress was reached at about 70d under normal condition. The power ramp can increase the cladding temperature and has visible influence on the stress distribution. The hoop stress of the cladding reversed under LOCA condition. The tensile hoop stress on the outer surface significantly increased, which caused the obvious increase of failure probability of monolithic SiC, and the failure probability of SiC layer is significantly increased. The conclusion of the analysis has guiding significance for the theoretical design of SiC composites.
Use of Micro-Gravity Sensors for External Fluid Level Monitoring in Waste and Nuclear Related Applications
Authors: Ridgway Jeff , Slater Bryant 
Published: October 2020

Abstract
There are a number of applications in nuclear energy and hazardous waste disposal that require monitoring of fluids under extreme environments, including high levels of temperature, pressure, toxicity and radioactivity. Many of these applications will benefit from a monitoring technique that is external and non-invasive. Currently the sensors used are invasive, must reside inside the pressurized vessels and must penetrate the vessel walls, which can create a weakness in the vessel. Additionally, instruments that are used inside such containers must be exceptionally hardened to the environment. Information Systems Laboratories (ISL) has developed an external mass (gravimetric) measuring technique for monitoring nuclear coolant in Small Modular Reactors (SMRs), which will also work for measuring fluid levels in waste tanks, that avoids the problems inherent in invasive sensors. It utilizes a COTS gravitational sensor of unprecedented accuracy, leveraged via proper sensor placement geometry, to detect fluid changes of small amplitude from an outside position, obviating the need to penetrate the vessel. The technique is called Gravisense™.
ISL has proven via simulation and experiment that this concept can be usefully applied to monitoring fluid levels in both nuclear reactors and large waste tanks. Numerical simulation algorithms were developed to calculate the gravity effect of small changes in water level, which were verified by experiments at the NIST Physical Simulator facility at the Oregon State University. The measured ultra-low noise levels of the superconducting gravimeter type which utilizes a Niobium sphere suspended in a magnetic field to attain its phenomenal accuracy, demonstrated that fluid levels in SMRs can be measured at least to within 3 cm. Furthermore, the method can distinguish between a contained leak (from reactor to containment vessel) from an external leak (from reactor to outside of containment).
Additionally, simulations of waste canisters that hold spent fuel rods show that the fluid level measuring accuracy can potentially do better than 1 cm accuracy by measuring from below the vessel, and judicious placement of sensors on top of large waste tanks can potentially achieve a very impressive 2 mm measurement accuracy.
These encouraging results prove that the Gravisense™ technique for fluid determination can be very useful in nuclear energy generation, testing, and research, as well as in waste monitoring situations that are difficult to monitor via traditional sensing technology. We believe that the next step should be to test the technique on canisters of the type that are currently storing waste in various DOE locations.
A Numerical Simulation Method for Core Internals Behavior in Severe Accident Conditions: Chemical Reaction Analyses in Core Structures by JUPITER
Authors: Yamashita Susumu , Kino Chiaki , Yoshida Hiroyuki 
Published: October 2020

Abstract
In order to contribute the improvement of estimation accuracy for severe accident code such as SAMPSON, we have developed the chemical reaction model such as eutectic reaction and oxidation in micro scale, e.g., B4C-SUS in the control rod blade and UO2-Zry in fuel rods, and implemented them to the computational fluid dynamics code named JUPITER. And we try to develop the coupled analysis frame work using SAMPSON and JUPITER to decrease uncertainty due to micro scale phenomena which cannot be calculate by severe accident analysis codes.
From the preliminary analysis in fuel rod heating analysis by JUPITER using SAMPSON output data, it was revealed that the implemented chemical reaction models work stably and obtain reasonable results.
Sensitivity Analysis for Release and Transport Behavior of Radionuclide of the 1F Unit-1 Accident Using SAMPSON
Authors: Kino Chiaki , Karasawa Hidetoshi , Uchida Shunsuke 
Published: October 2020

Abstract
To support the 1F decommissioning project, the Institute of Applied Energy has been analyzing the course of the accident using the SAMPSON code with an aim to grasp radionuclide distribution in 1F NPP. The present study has mainly focused on molybdenum release from a fuel pellet and chemical forms of iodine during the early phase of the accident because a release rate of semi-volatile FPs has various uncertainties. The present simulation set the detrimental factor of molybdenum used in FP release model as a sensitive parameter. The simulation results indicate that more molybdenum release causes more generation of hydrogen iodide and more deposition of iodine inside Reactor Pressure Vessel (RPV) because the adsorption velocity of hydrogen iodide is comparatively higher than deposition velocities of another FPs.
Development of Effective Momentum Model for Steam Injection Through Multi-Hole Spargers Using a Condensation Region Approach
Authors: Wang Xicheng , Grishchenko Dmitry , Kudinov Pavel 
Published: October 2020

Abstract
The Pressure Suppression Pool (PSP) in a Boiling Water Reactor (BWR) is served as a heat sink to prevent containment over-pressure. The steam can be injected through the multi-hole spargers. The development of thermal stratification where a thermocline with a large temperature gradient appears in the pool can lead to the higher pressure in the dry well compared with completely mixing pool conditions. Prediction of the thermal phenomenon in the pool is necessary for the support of system design and operation. Thus, the Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed. The models can be applied to CFD code by using (i) source terms in the transport equations or (ii) using respective boundary conditions at the Steam Condensation Region (SCR). Previous validation against PPOOLEX and PANDA tests using the source terms approach faced challenges in momentum distribution. Therefore, a preliminary investigation of using the second method was performed. The encouraging results implied that it is possible to further develop this approach. The goal of this work is to further develop the EHS/EMS models for the steam injection through a multi-hole sparger through the SCR model (i.e approach (ii)) and to validate it against the experimental data obtained from PANDA HP5 tests. Modeling guidelines are proposed. The temperature evolutions and vertical velocity profiles of these tests are compared to the simulation results. The agreement suggests that this model can provide an adequate estimation of the pool behavior.
Analytical Study of Perforation Damage to Reinforced Concrete Slabs Subjected to Oblique Impact by Projectiles With Different Nose Shapes
Authors: Kang Zuoyi , Okuda Yukihiko , Nishida Akemi , Tsubota Haruji , Li Yinsheng 
Published: October 2020

Abstract
Considerable research has been carried out to establish a rational assessment method for nuclear power plants against local damage caused by an accidental projectile impact. Most of the empirical formulas that have been proposed seek to quantitatively investigate local damage to reinforced concrete (RC) structures caused by a rigid projectile impact. These formulas have been derived based on impact tests performed perpendicular to the target structure, while only a few impact tests oblique to the target structure have been studied. This study aims to propose a new formula for evaluating local damage to RC structures caused by oblique impact based on experimental and simulation results. At present, we have validated an analytical method via comparison with experimental results and have conducted simulation analyses of oblique impact assessments on RC slabs using various projectiles with flat nose shapes. In this study, the same analytical method will be used to investigate perforation damage to RC slabs subjected to oblique impact by projectiles with hemispherical nose shapes. In this paper, the effects of projectiles’ nose shapes on perforation damage to RC slabs, the residual velocity of projectiles and the time history of energy transmission will be discussed.
Applicability of FLASH-CAT Model to Cable Tray Fire Modeling in Zone Model BRI2002
Authors: Shirai Koji , Tasaka Koji , Udagawa Toshiko 
Published: October 2020

Abstract
To clarify the heat and smoke propagation in multi-compartments under the spread of cable fire, a large-scale multi-compartment fire test (hereinafter the CFS-2 test) was performed by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) in France within the framework promoted by the Nuclear Energy Agency (NEA) in Organization for Economic Co-operation and Development (OECD) program PRISME2 (OECD/NEA, 2017). In the CFS-2 test, two rooms of a large-scale facility were adopted and these rooms have an identical volume (120 m3) enclosed with fire walls and were connected by a doorway (0.8 m in width and 2.17 m in height). As a fire source, five-layer cable trays (tray length of 2.4m, tray width of 0.45m and separation distance between trays of 0.3 m) with a fire-retardant PVC cable (77 kg) were used and ignited by a propane gas burner. The power level of the propane gas burner was set to around 80 kW. Moreover, all rooms were mechanically ventilated, and the renewal rate was 15 times per hour (3600 m3/h). During the fire test, the mass loss rate of fuel, gas and soot mass concentration, gas temperature, and etc. were measured. The measured peak values of the HRR, the mass loss rate and gas temperature were about 800 kW, 58 g/s and greater than 600 °C, respectively (Zavaleta, 2017).
As a fire model predicting fire characteristics in a compartment, a two-zone model, which divides the fire room into the hot smoke upper layer and lower layer consisting of cool fresh air, is widely used due to the advantages of the brevity of the calculation routine and the reliability of the calculation results. Among them, the BRI2 series, developed in Japan, is now reaching the current BRI2002 software (Wakamatsu, 2004) after several upgrades to improve the calculation precision. The Central Research Institute of Electric Power Industry (CRIEPI) introduced the cable tray fire source model based on the FLASH-CAT (Flame Spread over Horizontal Cable Trays) developed by National Institute of Standards and Technology (NIST) (McGrattan, 2012) into the zone code BRI2002. By comparing the numerical results with the experimental values measured during the CFS-2 test, the methodology for ignition time delay of each tray and horizontal flame propagation speed for each tray were discussed.
Optimization Study for Uncertainty Reduction of In-Pile CHF Experiments in TREAT
Authors: Seo Seokbin , Brown Nicholas R. , Armstrong Robert J. , Folsom Charles P. , Jensen Colby B. 
Published: October 2020

Abstract
Reactivity-initiated accidents (RIAs) are one of the postulated incidents that can threaten the operational safety of a nuclear reactor. During a RIA, a rapid increase of energy deposition in the fuel can lead to a departure from nucleate boiling (DNB) occurrence which refers to the point where a drastic decrease in heat transfer capabilities occurs and the surface heat flux exceeds the critical heat flux (CHF). Aiming to understand the fundamentals beneath CHF and to predict it, the Transient Reactor Test (TREAT) facility at the Idaho National Laboratory (INL) is a unique facility that will be used to experimentally investigate the transient CHF under in-pile pool boiling condition. As part of a comprehensive effort to utilize TREAT for this project, this study analyzed the expected uncertainties in the experimental data by identifying the key inputs for the uncertainty in the temperature measurements and quantifying their priorities. The sensitivities of key inputs from neutronics modeling, the clad-to-coolant heat transfer, thermophysical properties of the tube, and coolant conditions were quantified using Sobol sensitivity analysis methods, and the significant effect of the occurrence of the CHF on the sensitivity of input was found.
Development and Assessment of a Nearly Autonomous Management and Control System During a Single Loss of Flow Accident
Authors: Lin Linyu , Athe Paridhi , Rouxelin Pascal , Dinh Nam , Lane Jeffrey 
Published: October 2020

Abstract
In this work, a Nearly Autonomous Management and Control (NAMAC) system is designed to diagnose the reactor state and provide recommendations to the operator for maintaining the safety and performance of the reactor. A three layer-hierarchical workflow is suggested to guide the design and development of the NAMAC system. The three layers in this workflow corresponds to knowledge base, digital twin developmental layer (for different NAMAC functions), and NAMAC operational layer. Digital twin in NAMAC is described as knowledge acquisition system to support different autonomous control functions. Therefore, based on the knowledge base, a set of digital twin models is trained to determine the plant state, predict behavior of physical components or systems, and rank available control options. The trained digital twin models are assembled according to NAMAC operational workflow to support decision-making process in selecting the optimal control actions during an accident scenario.
To demonstrate the capability of the NAMAC system, a case study is designed, where a baseline NAMAC is implemented for operating a simulator of the Experimental Breeder Reactor II (EBR-II) during a single loss of flow accident. Training database for development of digital twin models is obtained by sampling the control parameters in the GOTHIC data generation engine. After the training and testing, the digital twins are assembled into a NAMAC system according to the operational workflow. This NAMAC system is coupled with the GOTHIC plant simulator, and a confusion matrix is generated to illustrate the accuracy and robustness of implemented NAMAC system. It is found that within the training databases, NAMAC can make reasonable recommendations with zero confusion rate. However, when the scenario is beyond the training cases, the confusion rate increases, especially when the scenarios are more severe. Therefore, a discrepancy checker is added to detect unexpected reactor states and alert operators for safety-minded actions.
Study on Seismic Isolation and Hi-Frequency Vibration Isolation Technology for Equipment in Nuclear Power Plant Using Aero Floating Technique
Authors: Takito Kiyotaka , Furuya Osamu , Kurabayashi Hiroshi , Sanpei Kunio 
Published: October 2020

Abstract
In Japan, most structures on the ground surface need seismic countermeasures because of frequently earthquakes. On the other hand, vibration isolation devices are applied to precision or important equipment in several facilities that dislikes vibration in order to reduce daily vibration. In general, vibration isolation devices are intended for high frequency and small amplitude range. However, it is difficult to cut off both vibration region caused by flying object collision and seismic motion with existing technologies. The authors propose insulation of equipment and vibration transmitted through the floor by floating equipment, and have. We have devised and built an air floating device that operates when a trigger input is applied to save the energy of this dynamically acting device. It was estimated by numerical calculation that the aero floating device keeps lifting stably in the condition with the air pressure in the auxiliary air chamber about 75 to 80 kPa. The performance specifications of the proposed device were verified from shaking table test.
As a result, the effect of reducing the maximum acceleration by about 1/5 against the seismic motion of El Centro NS, Taft NS, Tohoku NS, and Hachinohe EW was confirmed by floating the mass on the frame assuming the equipment. From the obtained power spectrum diagram (PSD) of the response acceleration, it was confirmed that all frequency components up to 25 Hz is reduced by using proposed aero floating base isolation device.
Experimental Study for Evaluation of Spreading Behavior of Free-Falling Molten Core With Floor Impingement
Authors: Ito Tomomasa , Watanabe Yuta , Yamamoto Yasunori , Sahboun Nassim , Miwa Shuichiro 
Published: October 2020

Abstract
Understanding the spreading behavior of the molten core is important for predicting the progress of severe accidents and for smooth decommissioning work. In this study, we performed molten metal drop experiments using Sn and Cu to obtain fundamental data and its uncertainties that can be used to verify the analysis model and to evaluate the effect of scaling parameters on the spreading and deposition behaviors. We obtained spreading area and deposition thickness of the metals after floor impingement. The standard deviation on average were 6.0% and 6.7% for the thickness and the area, respectively. In addition, the correlation with each dimensionless number was confirmed, and strong correlations were obtained between the deposition behavior and the Re number, and between the spreading behavior and the Pe number. We constructed a correlation equation using our experimental data as a function of Re number, We number, Fr number, Pe number, and dimensionless length (L/d). The errors for our experimental data were within 30%.
Fundamental Study on Decontamination Method Using Shot Blasting and Barrel Polishing for Radioactive Metal Waste
Authors: Yamamoto Sota , Takahashi Hideharu , Taniguchi Hayato , Kawashima Akihiko , Takahashi Hiroshi , Kikura Hiroshige 
Published: October 2020

Abstract
In the decommissioning of nuclear power plants (NPP), we can greatly reduce the amount of low-level waste (LLW) by decontamination process. Generally, chemical decontamination method that use chemical solutions to dissolve the deposited waste is major way of decontamination. But this method takes a lot of time and generates secondary waste. On the other hand, the mechanical decontamination method that projected particles such as metal to decontamination target and remove the deposited waste takes shorter time than chemical decontamination method usually. However, in the case of mechanical decontamination of the small diameter pipe to be used in the primary system of the nuclear reactor is difficult because we must divide it vertically into half, and these small diameter pipes are disposed as LLW. At the decommissioning of advanced thermal reactor Fugen (operation finished in 2003) in Japan, the new decontamination machine which compounded with the blasting machine which projected particles such as metal and the barrel polishing machine which is a machine that polishes like a washing machine using particles such as metal was produced experimentally for the pipes of such a small diameter. By using this machine, the radioactivity of contaminated undivided pipe is decreased, and we found that this machine can use as a decontamination machine. In this study, we operated the blasting and barrel grinder independently in order to elucidate the mechanism of the decontamination system. We placed flat metal simulators in each device and measured mass changes in each to compare which decontamination effect was the primary one. The results suggest that the decontamination effect of the blasting machine is higher than that of the barrel polishing machine. We also used a high-speed camera to observe the decontamination of the blasting system and to investigate the interaction between the pipe sample and the particles. It was suggested that the number of particles entering the inside of the pipe increases as the pipe faces the projection direction.