bluuebunny/crossref_metadata_2025_split huggingface dataset:
Development of Ultrasonic Measurement System for Shape and 2D Velocity Field Using Ultrasonic Velocity Profiler and Total Focusing Methods
Authors: Zhang Zeliang , Liu Tianrun , Batsaikhan Munkhbat , Takahashi Hideharu , Kikura Hiroshige 
Published: October 2021

Abstract
A severe accident at TEPCO’s Fukushima Dai-ichi Nuclear Power Plant (FDNPP) occurred due to a strong earthquake and massive tsunami that struck Japan’s eastern region on 11 March 2011. During this accident, the reactor core of Units 1, 2, and 3 of the 1F melted and fuel debris formed in the reactor pressure vessel (RPV) and primary containment vessel (PCV). The Japanese Government has drawn a mid- and long-term roadmap towards 1F decommissioning. However, three reactors’ decommissioning work has faced difficulties due to the lack of realistic information about the damaged cores such as the distribution of fuel debris. Since the radiation levels inside the reactor buildings have been too high for human access, the robots and cameras have been using to investigate the internal status of PCV. To accelerate the decommissioning process of 1F, determining of fuel debris information and contaminated water leakage is crucial to decide further decommissioning steps and strategies. Therefore, in this study, we are developing an ultrasonic measurement system that can obtain object’s shape and the velocity around it. Because ultrasonic transducers are robust against high radiation dose and applicable in opaque liquids. The ultrasonic system used in the present study consists of two sectorial array transducers, which can measure object shape and 2D vector flow field. To measure the object’s velocity and shape that represents fuel debris, Ultrasonic Velocity Profiler (UVP) and Total Focusing Methods (TFM) are employed, respectively. Based on measurement results of the 2D vector field and shape of the object, we can predict contaminated water leakage and distribution of object at the bottom of the PCV in 1F.
Research on Start-Up Design of Nuclear Safety Level Parallel Redundant Control Station
Authors: Shi Guilian , Shou Yunxu , Li Gang 
Published: October 2021

Abstract
Compared with the operating life cycle of the digital I&C system in nuclear power plants, the start-up process of the control station is minimal and easily overlooked. A design that is too simple is not suitable for nuclear power applications. The complexity of the start-up design comes from three aspects: One is the diversity of start-up scenarios. In addition to the start of the normal plan, there are unexpected start-ups that cannot be ignored; the second is the complexity of data synchronization in the redundant system; the third is the consideration of human factors. The start-up process involves a lot of human-computer interaction, and how to reduce human risk is also an important design requirement. If the factors are not considered properly, the control station will easily cause disturbance of the controlled equipment when starting, and may even cause the malfunction of the engineered safety features actuation system. This article focuses on the nuclear safety-level parallel redundant control station, analyzes various scenarios of the control station start-up, and synthesizes the design requirements for the start-up phase. According to the requirements, the overall design plan of “initialization-synchronization-comparison-commissioning” is proposed, and the human operation risks involved in each stage are analyzed, and corresponding prevention plans are proposed. The FirmSys parallel redundant control station implemented according to this scheme has been successfully applied in ten commercial nuclear power units including Unit 5 and Unit 6 of Yangjiang Nuclear Power Plant.
Numerical Analysis of Pressurized Thermal Shock in Reactor Pressure Vessel
Authors: Zhang Yubin 
Published: October 2021

Abstract
One of the most important issues for the assessment of life time extension of a nuclear power plant is whether the Reactor Pressure Vessel (RPV) can maintain structural integrity under Pressurized Thermal Shock (PTS) phenomenon. The most critical scenario takes places during cold water injection in the cold legs after a postulated Loss of Coolant Accident (LOCA). In the present study, a deterministic method has been conducted. The thermal-hydraulic system analysis program CATHARE, the Computational Fluid Dynamics (CFD) software STAR-CCM+ and the mechanical structure analysis software ABAQUS are used to simulate the Small Break and Intermediate Break Loss of Coolant Accident (SB-LOCA and IB-LOCA). As a result, the SB-LOCA and IB-LOCA accident transients are both simulated and the thermal stratification during the injection is demonstrated. The analysis results give the lowest temperature of SB-LOCA under vertical injection. Because of the bigger injection flow rate, the thermal shock is severer in IB-LOCA than SB-LOCA. According to the mechanical structural analysis results from ABAQUS, the thermal stresses are mainly concentrated on the both sides of the bottom of the inlet nozzle and the beginning of the downcomer of Reactor Pressure Vessel (RPV).
Hybrid Improved Empirical Mode Decomposition and Artificial Neural Network Model for the Prediction of Critical Heat Flux (CHF)
Authors: Djeddou Messaoud , Zhao Xingang , Hameed Ibrahim A. , Rahmani Ahmed 
Published: October 2021

Abstract
Three Hybrid artificial neural network (ANN) models namely radial basis function (RBF), generalized regression neural networks (GRNN), and multi-layer perceptron (MLP) combined with empirical mode decomposition (EMD) are developed for CHF predictive modelling using CHF experimental databases.
First, the original experimental inputs data series are decomposed into several intrinsic mode functions (IMFs) and one residual by EMD, whose components are divided into high, medium and low components. The performance parameters of the hybrid models indicates that the root mean square error (RMSE) are 0.8831, 0.6522, and 0.4149; the mean absolute error (MAE) are 0.6697, 0.4636, and 0.1935. The values of the R-square of the developed prediction approach utilizing EMD-RBF, EMD-GRNN, and EMD-MLP models are 0.8553, 0.9302, and 0.9818, and the index of agreement are 0.9464, 0.9700, and 0.9894., The value of the R-square and the index of agreement of the proposed models are much higher than those of the simple models.
The Pearson’s test results show that the association strength between the measured and the predicted values of the proposed model EMD-MLP is the strongest. These results show the following: (a) compared with other related, recent studies, the prediction accuracy of the hybrid model EMD-MLP proposed in this research is the best hybrid model; (b) the proposed hybrid model (EMD-MLP) attains superior performance compared with simple models.
A Small PWR-Core Physical Calculation Based on PWR-Core Analysis Code CORAL
Authors: Yang Wen , Zhou Lun , Qiu Junrong , Tai Yun 
Published: October 2021

Abstract
Three dimensional PWR-core analysis code CORAL is developed by Wuhan Second Ship Design and Research Institute. This code provides basic functions including three-dimensional power distribution, fine power reconstruction, fuel temperature distribution, critical search, control rod worth, reactivity coefficients, burnup and nuclide density distribution, etc. CORAL employ nodal expansion method to solve neutron diffusion equation, and the least square method is used to achieve few group constants, and sub-channel model and one-dimensional heat transfer is used to calculate fuel temperature and coolant density distribution, and burnup distribution and nuclide nuclear density could be obtained by solving macro-depletion and micro-depletion equation. The CORAL code is convenient to update and maintain in consider of modular, object-oriented programming technology. In order to analyze the computational accuracy of the CORAL code in small PWR-core and its capability to deal with heterogeneous, calculation analysis are carried out based on the material and geometry parameters of the SMART core. The core has 57 fuel assemblies, with 8, 20 or 24 gadolinium rods arranged in the fuel assemblies. In this paper, a quantitative comparison and analysis of the small PWR problem calculation results are carried out. Numerical results, including effective multiplication factor, assembly power distribution and pin power distribution, all agree well with the calculation results of OpenMC or Bamboo at both hot zero-power (HZP) and hot full-power (HFP) conditions.
Research on Internal Fire Ignition Frequency of Fire Probability Safety Analysis in Small Module Reactor
Authors: Chen Yanzhu , Yang Zhichao 
Published: October 2021

Abstract
The equipment number of Small module reactor (SMR) is significantly less than that of commercial reactor, and accordingly, the number of fixed fire ignition sources and the ignition frequency are much less. Based on the analysis of the design characteristics of Small Module Reactor, this paper develops a preliminary research and innovation on the internal fire ignition frequency evaluation method of Small Module Reactor, and implements a sensitivity analysis on the commercial reactor ignition frequency methods. The result indicates the fire ignition frequency would increase 51% if the NUREG/CR-6850 generic ignition frequency data are adopted. However, the design characteristics of SMR can be better reflected in this innovative method, which needs more attention in the future.
Low Dose Assessment Uncertainty Analysis for Landauer® nanoDot™ OSLDs
Authors: Aras Egemen , Hayes Robert 
Published: October 2021

Abstract
This study aims to low-level background range measurements of commercial OSLDs. This work is the initial step to put a framework to detect any illicit nuclear activities in any nuclear facility at any time. The idea in this framework is to utilize already placed OSLDs in the facility which are normally read-out periodically. The results of these measurements could be distinguishable from the background radiation since this paper shows how the background dose with its statistical fluctuation provide detection limits in these applications. To do this, we measured dosimeters in two ways; without removal or replacement and full removal for each measurement. As a result of measurements, the initial dose, bleaching constant, and background dose was evaluated for different measurements. ANOVA was applied to all measurements and all measurements considered a measurement data set to analyze results. Consequently, we observed no statistically significant difference in these different kinds of measurement approaches relative to the total propagated uncertainty in any given dose estimate. This shows a passive detection can be verified with iterative measurements to improve statistics without compromising data quality when coupled with dose levels of potential interest serves to advance this potential nonproliferation application.
A Systematic Review of PCTRAN-Based Pressurized Water Reactor Transient Analysis
Authors: Racheal Suubi , Liu Yongkuo , Ernest Miyombo , Ayodeji Abiodun 
Published: October 2021

Abstract
The impact of nuclear accidents has been a topic of debate since the construction of the first nuclear reactor, and still stands as a key issue of public concern. Several codes and simulators have been used to study the transient progression in pressurized water reactors, and to evaluate the technical measures adopted to scale down the risk of accidents. However, some of these codes are not suitable for multipurpose research and training as they require significant user expertise, leading to analysis uncertainties largely from the code user effect. This paper presents a bird-eye view of one of the most widely used nuclear reactor transient analyzer — the Personal Computer Transient Analyzer (PCTRAN). This paper discusses the comparative advantages of the simulator from the users’ perspective, with specific attention to its utilization both for research and training. The paper also demonstrates the ease of usage by simulating common transient in a pressurized water reactor. Finally, observations and possible improvements to the code to increase its usability in research, education and training are discussed. This work aims to evaluate the robustness of the simulator towards better utilization for research and training, especially in nuclear newcomer countries.
Experimental Study on the Critical Heat Flux of the Zirconium Alloy Microstructure Surface Fabricated by Ultraviolet Laser
Authors: Ren Quan-yao , Wang Haoyu , Zhu Fawen , Li Yuanming , Zhang Lin , Pu Zengping , Yuan Pan , Ran Renjie , Huang Chunlan , Li Quan , Wang Xiaoliang , Liu Yongda , Xu Jie 
Published: October 2021

Abstract
The critical heat flux of fuel assembly was of great interest for pressurized water reactors, which was related to the safety and economic efficiency. Recently, many studies showed that the nano- and micro-structures on the heat surface could improve the CHF greatly by changing the wettability, onset of nucleate boiling, bubble behaviors (grow, slip, departure etc.) on the modifying surfaces. In present study, the square column microstructures was manufactured on the zirconium alloy surface with ultraviolet laser, whose column microstructure images acquired by SEM (scanning electron microscope) showed many smaller microstructures in the bottom region than the size of square column. And the modified surface showed anisotropic hydrophily, which was smaller than that on the original surface. And the CHF of zirconium alloy microstructure surface was improved about 25%, which was a considerable value for the fuel assembly. And more studies are recommended to focus on the stability and flow boiling behaviors of the zirconium alloy microstructure surface in the future study.
Mixing Characteristic Measurement of Flow in Reactor Pressure Vessel by Laser Induced Fluorescent Method
Authors: Chen Mingpeng , Xie Guanhui , Li Dongyang , Tan Sichao 
Published: October 2021

Abstract
Under normal conditions, the mixing uniformity of boric acid and coolant in the descending section and the lower chamber of PWR is an important factor affecting the core reactivity distribution. In the emergency shutdown process, the transient dilution of boron may cause secondary criticality. Therefore, the mixing of the de-borated slug with the ambient coolant in the reactor pressure vessel was investigated by using a three-loop 1:6 scaled visualization pressure vessel model with ACP1000 as the prototype device. The pressure vessel model retains the structures related to the experiment, such as the radiation monitoring tube, while the other structures are simplified accordingly. The dye solution was used as the boric acid injection coolant to measure the concentration distribution of dyes on the vertical and different cross sections, and the mixing mechanism was analyzed by LIF. In the slug mixing test, the coolant flow rate and the dye injection flow rate were changed respectively, and the chromatid concentration changes with time and its distribution at different times were detected. The effect of Reynolds number is analyzed. In the descending section, the injection continuously collides with the tube wall and is diluted by the surrounding pure water, the transport of the stain along the annulus boundary and the diffusion on the cross section are weakened, and the transport mainly occurs in the circular direction. After the mixed fluid passes through the lower head and the lower plate of the core, the concentration of the fluid region near the wall is lower, while the concentration in the middle is higher, and the distribution is inverted fan. The related research results of this experiment qualitatively reveal the mudding process of boric acid and deborated coolant before entering the reactor core, quantitatively analyzed the concentration distribution of the observed area.
Baseline WALT DNB Test Results With Cr-Coated Cladding to Support Accident Tolerant Fuel Development
Authors: Wang Guoqiang , Byers William A. , Karoutas Zeses 
Published: October 2021

Abstract
When departure from nucleate boiling (DNB) occurs, the critical heat flux (CHF) must be accurately determined for different conditions in order to maintain reasonable margin to CHF during the operation of a pressurized water reactor (PWR). Cladding surface texture may influence DNB by providing boiling nucleation sites and by introducing turbulence that inhibits film boiling. Textured surfaces can increase or decrease CHF, depending on their structure. The chemistry of the cladding surface controls wetting. Stronger wetting inhibits detachment of coolant from the surface and initiation of film boiling. Westinghouse has developed a new cladding product that consists of a metallic chromium layer on Optimized ZIRLO™ cladding. Since both the surface microstructure and chemistry of the Cr-coated cladding are different from the zirconium alloy surfaces that have been previously characterized, it is necessary to show that DNB margin is similar or greater than current cladding surface.
In this paper, the Westinghouse advanced loop tester (WALT) generated DNB test data at PWR thermal hydraulic and chemistry conditions to support the Westinghouse EnCore® accident tolerant fuel (ATF) products. With an Apple-to-Apple comparison DNB strategy, the Westinghouse WALT loop team has been performing DNB tests using heater rods with Cr-coated and Optimized ZIRLO uncoated surfaces. Meanwhile, the cladding local resistivity measurement is critical for an accurate calculation of the local heat flux. Under this defined DNB strategy, a new methodology has been developed for performing DNB tests and measuring/calculating the CHF values for different cases.
In this paper, the new methodology is described for local resistivity measurement and heat flux calculations. In addition to measuring the CHF for different coolant conditions, possible cladding failure modes were also investigated during the DNB tests. There has been concerns that the coating would spall off from cladding surface, or cracking of the cladding surface might be enhanced. The cladding section around the DNB location were examined to address such concerns.
Westinghouse eVinci™ Heat Pipe Micro Reactor Technology Development
Authors: Swartz Matt M. , Byers William A. , Lojek John , Blunt Rory 
Published: October 2021

Abstract
Westinghouse Electric Co. is developing the eVinci™ Micro-Reactor to provide grid resiliency for strategic infrastructure, deliver power for remote applications and support growing space fission power needs. At the heart of the reactor technology are alkali metal heat pipes. Alkali metal heat pipes replace the complex primary coolant systems necessary for other high temperature reactor designs with a safe, simple system that has no moving parts. The heat pipe reactor relies on alkali metal phase change resulting in superior temperature uniformity within the reactor core and elimination of the need for a pressurized primary system containment vessel. An array of closed heat pipes provides redundancy of the primary heat removal path and enables a modular core design that supports reduced scale testing.
Westinghouse has commissioned a dedicated facility aimed at developing heat pipe technology for commercial nuclear applications. The efforts build upon the success of the KRUSTY nuclear demonstration and solves some of the challenges associated with commercial micro-reactors that require use of low enriched uranium (LEU). Westinghouse has completed first of a kind demonstration testing of high-performance alkali metal heat pipes using Iron, Chromium, Aluminum (FeCrAl) alloys. This technology enables a LEU micro-reactor core in addition to an open-air Brayton thermodynamic power conversion cycle.
Structural Integrity Assessment of a Unit Cell in a Laboratory-Scale Printed Circuit Heat Exchanger for Molten Salt Reactors With Supercritical CO2 Power Cycle
Authors: Che Shuai , Zhang Sheng , Burak Adam , Sun Xiaodong 
Published: October 2021

Abstract
The Printed Circuit Heat Exchanger (PCHE) is considered promising as an intermediate heat exchanger for Molten Salt Reactors (MSRs) due to its highly compact construction, high heat transfer effectiveness, and capability of withstanding high pressures. In this study, thermal-mechanical simulations were performed using a two-channel unit-cell model with the attempt to investigate the structural integrity of a laboratory-scale PCHE that was designed for molten salt-to-supercritical carbon dioxide heat transfer, with the temperature field obtained from Computational Fluid Dynamics (CFD) simulations. It is shown that the fillets on the semi-circular channel walls are stress concentration regions and that the stress intensity decreases quickly as the distance from the fillets increases. A quick drop in the maximum stress intensity is observed with the increase of the fillet radius. There is no significant increase in the stress intensity for locations around the zigzag sharp corners. With a lower bulk temperature and a higher stress intensity, the region close to the outlet of the PCHE hot channels is more vulnerable to potential failures than the inlet region of the hot channels. In addition, the choice of channel models has a weak impact on the maximum stress intensity around the cold channel fillets.
Strategies for Application of Reliability and Integrity Management for Inservice Inspection of Compact Heat Exchangers in High Temperature Reactors
Authors: Keating Robert , McKillop Suzanne , Jentz Ian 
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 exchangers manufacturers CompRex and Vacuum Process Engineering.
The objective of the research was to enable the use of compact heat exchanger designs in high temperature advanced reactor service in order to improve plant efficiency and economics. A necessary step for achieving this objective is to ensure that the ASME Boiler and Pressure Vessel Code, Section III has rules for the construction of CHXs for nuclear service. However, construction rules alone are not sufficient to deploy a compact heat exchanger in an advanced reactor. ASME Section XI Rules for the Inservice Inspection (ISI) of a heat exchanger in an operating nuclear reactor will be required, and mostly likely ISI will be conducted in according with ASME Section XI, Division 2, Reliability and Integrity Management (RIM) methodology, which is a non-deterministic, technology neutral, approach to inservice inspection.
Since a RIM based ISI program is developed by an owner considering the entire reactor technology safety case, owners will need specific information regarding the performance, failure modes, reliability, and inspection options for compact heat exchangers. Previously, ICONE27-1139 provided strategies for inservice inspection of compact heat exchanger that outlined how an owner can use the planned IRP research to support the development of an ISI program that includes compact heat exchangers.
Now that the IRP research is completed, this paper updates the earlier strategy to reflect the results of the research. Failure mechanisms for compact heat exchanger are unique and not widely understood. The IRP has studied the failure mechanisms, monitoring and inspection technologies and this research may be used to enable an owner/operator of an advanced nuclear reactor to develop a RIM based inservice program.