bluuebunny/crossref_metadata_2025 huggingface dataset:
Characteristics and proposed mechanisms of QRS morphology observed during the left bundle branch pacing procedure
Authors: Shimeno Kenji , Tamura Shota , Nakatsuji Kenichi , Hayashi Yusuke , Abe Yukio , Naruko Takahiko
Published: October 2021

Background

In performing left bundle branch pacing (LBBP), various QRS morphologies are observed as the lead penetrates the ventricular septum (VS). This study aimed to evaluate these characteristics and infer the mechanism underlying each QRS morphology.

Methods

In 19 patients who met the strict criteria for LBB capture, we classified the QRS morphologies observed during the LBBP procedure into seven patterns, the first five of which were determined by the depth of penetration: right ventricular septal pacing (RVSP), intraventricular septal pacing (IVSP1 and IVSP2), endocardial side of left ventricular septal pacing (LVSeP), nonselective LBBP (NS‐LBBP), selective LBBP (S‐LBBP), and NS‐LBBP with anodal capture. The parameters of the QRS morphologies in these seven patterns were evaluated.

Results

Among the first five patterns, stimulus‐QRSend duration (s‐QRSend) was the narrowest in IVSP1 rather than in NS‐LBBP, and stimulus‐to‐peak of R wave in V6 (s‐LVAT) was significantly shortened in two steps, from RVSP to IVSP1 (96 ± 11; 82 ± 8 ms, p < .01) and from LVSeP to NS‐LBBP (76 ± 7; 60 ± 4 ms, p < .01). The late‐R duration in V1 was significantly prolonged in the order of LVSeP, NS‐LBBP, and S‐LBBP (45 ± 7; 53 ± 10; 71 ± 15 ms, respectively, p < .01).

Conclusions

s‐QRSend was the narrowest in IVSP1 rather than in NS‐LBBP among the QRS morphologies observed during lead penetration through the VS. The prolonged late‐R duration in V1 and abrupt shortening of the s‐LVAT in V6 may help determine LBB capture during lead penetration.
The association between perioperative pupillary parameters and postoperative acute pain: A pilot cross‐sectional study
Authors: Fang Pan‐pan , Wu Jun‐qi , Tang Li‐li , Gao Shan , Liu Xue‐Sheng
Published: October 2021

Objective

We aim to explore the capacity of perioperative pupillary variables to predict acute pain in the post‐anesthesia care unit (PACU).

Methods

Patients scheduled to undergo thoracic or abdominal surgery under general anesthesia between April 2021 and June 2021 were enrolled. We measured the pupil diameter, pupillary light reflex (PLR), and pupillary reflex dilatation 5 min before anesthesia induction (T1), 5 min after intubation (T2), at the end of anesthesia (T3), immediately before extubation (T4), and 5 min after extubation (T5). We assessed the early postoperative pain intensity in the PACU using Numeric Rating Scales (NRS) at recovery, 5 min after recovery, and 10 min after recovery. Logistic regression models were used to evaluate the association between perioperative pupillary variables and postoperative pain intensity.

Results

Fifty‐one patients were enrolled, 50 of whom were included in the final analysis. A total of 13 patients (26%) needed remedial analgesia in the PACU. Pupil parameters at T1, T2, T3, and T5 were not associated with NRS in the PACU. Multiple logistic regression models and receiver operating characteristic (ROC) curves indicated that only latency of PLR at T4 can predict postoperative acute pain. The ROC analysis showed that the cutoff value for latency of PLR at T4 was 0.29 s to discriminate between no pain and pain, and the area under the curve was 0.778 (95% CI 0.634–0.922, p = 0.002) with sensitivity 50.0% and specificity 91.7%.

Conclusion

The latency of PLR immediately before extubation may be a useful predictor for postoperative acute pain in the PACU.
Genome‐wide association study reveals that the cupin domain protein OsCDP3.10 regulates seed vigour in rice
Authors: Peng Liling , Sun Shan , Yang Bin , Zhao Jia , Li Wenjun , Huang Zhibo , Li Ziyin , He Yongqi , Wang Zhoufei
Published: October 2021

Seed vigour is an imperative trait for the direct seeding of rice. In this study, we examined the genetic regulation of seedling percentage at the early germination using a genome‐wide association study in rice. One major quantitative trait loci qSP3 for seedling percentage was identified, and the candidate gene was validated as qSP3, encoding a cupin domain protein OsCDP3.10 for the synthesis of 52 kDa globulin. Disruption of this gene in Oscdp3.10 mutants reduced the seed vigour, including the germination potential and seedling percentage, at the early germination in rice. The lacking accumulation of 52 kDa globulin was observed in the mature grains of the Oscdp3.10 mutants. The significantly lower amino acid contents were observed in the mature grains and the early germinating seeds of the Oscdp3.10 mutants compared with those of wild‐type. Rice OsCDP3.10 regulated seed vigour mainly via modulating the amino acids e.g. Met, Glu, His, and Tyr that contribute to hydrogen peroxide (H2O2) accumulation in the germinating seeds. These results provide important insights into the application of seed priming with the amino acids and the selection of OsCDP3.10 to improve seed vigour in rice.
Intelligent Model for Predicting Downhole Vibrations Using Surface Drilling Data During Horizontal Drilling
Authors: Saadeldin Ramy , Gamal Hany , Elkatatny Salaheldin , Abdulraheem Abdulazeez
Published: October 2021

Drillstring vibration is a major concern during drilling wellbore, and it can be split into three types: axial, torsional, and lateral. Many problems associate with the high drillstring vibrations as tear and wear in downhole tools, inefficient drilling performance, loss of mechanical energy, and hole wash-out. The high cost for the downhole measurement of the drillstring vibrations encourages machine learning applications toward downhole vibration prediction during drilling. Consequently, the objective of this paper is to develop an artificial neural network (ANN) model for predicting the drillstring vibration while drilling a horizontal section. The ANN model uses the surface drilling parameters as model inputs to predict the three types of drillstring vibrations. These surface drilling parameters are flowrate, mud pumping pressure, surface rotating speed, top drive torque, weight on bit, and rate of penetration. The study utilized a data set of 13,927 measurements from a horizontal well that was used to train the ANN model. In addition, a different data set (9284 measurements) was employed to validate the developed ANN model. Correlation coefficient (R) and average absolute percentage error (AAPE) are statistical metrics that are used to evaluate the model accuracy based on the difference between the actual and predicted values for the axial, torsional, and lateral vibrations. The results of the optimized parameters for the developed model showed a high correlation coefficient between the predicted and the actual drillstring vibrations that showed R higher than 0.95 and AAPE below 3.5% for all phases of model training, testing, and validation. The developed model proposed a model-based equation for real-time estimation for the downhole vibrations.
Manufacturability Estimation on Burnable Poison Mixed Fuel For Improving Criticality Safety of HTGR Fuel Fabrication
Authors: Hasegawa Toshinari , Fukaya Yuji , Ueta Shohei , Goto Minoru
Published: October 2021

Manufacturability of the BP-mixed uranium dioxide (UO2) kernel for burnable poison (BP) credit concept was estimated to ensure criticality safety in 14 wt% enrichment fuel fabrication facility for commercial high temperature gas-cooled reactor (HTGR). The case of the coating process has not been examined, so this study is different from manufacturability estimation on Tri-ISOtropic (TRISO) coated fuel. As a BP, boron, gadolinium, erbium, and hafnium were investigated. It is essential that mixing BP does not affect the shape of the fuel kernel, and that BP does not melt or vaporize during the heat treatment in the reduction process. We determined to add BP as a nitrate by considering the existing kernel fabrication process. The results of thermodynamic equilibrium analysis showed boric acid would melt and vaporize during the heat treatment processes. On the other hand, it was found that gadolinium, erbium, and hafnium nitrate would change to solid oxides that do not melt even at 2000 °C. These results indicate the existing external sol-gel method could be applied for the fabrication of the UO2 fuel kernel mixed with gadolinium, erbium, or hafnium.
Flow Induced Vibration and Fretting Wear Characteristics of Fuel Rods
Authors: Feng Zhipeng , Qi Huanhuan , Huang Xuan , Chen Guo , Liu Shuai , Zhang Yixiong
Published: October 2021

Fuel assembly is located in the reactor core, which is the key component of nuclear power plant reactor. Its structural integrity will directly affect the safety and reliability of nuclear reactor operation. Fretting wear between fuel rod and grid supports caused by flow induced vibration is currently one of the main causes of fuel rod failure, which is also the most concerned phenomenon in fuel assembly design. Based on the random vibration theory and considering the interaction between fuel rod and support grid, a theoretical model and evaluation criteria for response prediction are proposed for the three possible flow induced vibration mechanisms in Pressurized Water Reactor (PWR) fuel assemblies, namely, turbulent excitation, fluidelastic instability and vortex shedding. At the same time, based on the Archard’s equation, the fretting wear analysis model of fuel rods is established, which is suitable for industrial application. The effects of transverse velocity, axial velocity and fluid density on the response of flow induced vibration and fretting wear of fuel rods are studied. The results show that the amplitude/wear depth varies with the transverse velocity in a quadratic way, exponentially with the axial velocity, and linearly with the fluid density. The established analysis process and evaluation method can provide technical support for the design and optimization of fuel assembly.
Core Thermal-Hydraulic Analysis During Dipped-Type Direct Heat Exchanger Operation in Natural Circulation Conditions
Authors: Hamase Erina , Doda Norihiro , Ono Ayako , Tanaka Masaaki , Miyake Yasuhiro , Imai Yasutomo
Published: October 2021

In order to enhance the safety of sodium-cooled fast reactors, a direct reactor auxiliary cooling system (DRACS) under natural circulation conditions with a dipped-type direct heat exchanger (D-DHX) in an upper plenum of the reactor vessel has been investigated. During the DRACS operation, the complicated thermal-hydraulic phenomena that cold coolant from D-DHX flowed into the fuel subassemblies and narrow gaps between them, which is well-known as inter-wrapper flow (IWF) was observed. Therefore, a multi-dimensional thermal-hydraulic analysis model in the reactor vessel for computational fluid dynamics (CFD) code (RV-CFD model) has been developed to evaluate the core cooling performance under natural circulation conditions. For the design study, the RV-CFD model is demanded to simulate with reasonable calculation costs while maintaining accuracy. In this paper, the application of the subchannel analysis method by CFD code for fuel subassemblies (subchannel CFD model) to the RV-CFD model was attempted. In the subchannel CFD model, the porous media approach was used to consider local geometry in the fuel subassembly, and the effective heat conductivity coefficients in diffusion term of the energy equation were set to fit the actual radial thermal diffusion between subchannels. Analysis results were compared to the experimental data obtained in the sodium experimental apparatus PLANDTL-1 and the calculated sodium temperature in the core had good agreement with the experimental result. It was confirmed that the RV-CFD model with subchannel CFD model was applicable to the core thermal-hydraulic analysis during the DRACS with the D-DHX operation under natural circulation conditions.
The Development of the NPP Nuclear Emergency Drilling Assistant System
Authors: Yanfang Chen , Xueyan Hou , Fei Chao , Longze Li , Chuan He , Wen Yang
Published: October 2021

The study carries out the rapid simulation calculation by using the integrated severe accident analysis program MAAP, performs the 3D modeling of nuclear power plant with 3DMax software, and implements the effect rendering with Unity3d. A set of nuclear emergency drilling assistant system has been developed to meet the requirements of nuclear emergency drilling scripting and the decision-making support for emergency drilling in emergency headquarters. It realizes the super real-time simulation calculation of nuclear accident, generation and storage of nuclear emergency drilling scenario database, generation of nuclear emergency drilling script information, and real-time scenario reconstruction. The application of the system can effectively reduce the difficulty and workload of nuclear emergency drilling scripting, and improve the efficiency and credibility of the scripting. Additionally, it can vividly present the main phenomena and processes of nuclear accident, and provide the real-time dynamic data under the whole range and all working conditions for nuclear emergency drilling.
Key Element Analysis and Suggestion for Strengthen the Quality Management of Nuclear Power Plant Fasteners
Authors: Lu Yan , Ling Ligong , Xu Yu , Gao Chen
Published: October 2021

This article uses the letters and notices issued by the China National Nuclear Safety Administration on fastener quality incidents as the main line to review the process of fastener quality incidents from clues, investigation to handling. Based on the problems encountered in the process of handling this incident, the improvement direction and improvement measures for strengthening the quality management of fasteners were analyzed from the elements of quality assurance, design, procurement, manufacturing, installation, non-conformance management, inspection and testing, and documents. The design of nuclear power plants should clarify the classification and marking of fasteners, and the design of fasteners should be as universal and standardized as possible, and specifications and models should be reduced. Procurement should focus on supplier evaluation, limit the lowest bidding price, prevent cutting corners, etc.; manufacturing should set up witness points for key links, inspect the quality assurance system while witnessing; pay attention to the factory inspection of fasteners, and it is recommended to increase the re-inspection of fasteners before storage or use. Fasteners managed by equipment category are not recommended to be classified as bulk materials, but they still need to be re-inspected. Fastener re-inspection standards should take into account the requirements of different reactor types and different specification systems. Re-inspection should not only focus on mechanical properties, but also pay attention to the re-inspection of fastener surface quality. The re-inspection unit shall have a certain degree of credibility or be bound by nuclear safety regulations. The idea of establishing a professional nuclear industry fastener supply and testing platform was proposed.
The Development Status of Decommissioning Technology of Nuclear Facilities: An Insight From Patents
Authors: Zhang Yading , Mo Dan , Su Ran , Zhong Haoliang
Published: October 2021

The decommissioning technology of nuclear facilities has been one of the most important research directions of nuclear industry for several decades, and further enhancements are being carried out in the world. The global published patents about the decommissioning technology of nuclear facilities are investigated in this work. The technology forecasting and the development strategy of the decommissioning technology of nuclear facilities was provided from the view of patents, based on the nations, the technology field, the trend in the number of patents applications. This work will contribute to understanding the advanced developments of the decommissioning technology of nuclear facilities by a new perspective. Through quantitative statistics and content analysis, the understanding of relevant industries and technology development trend of personnel in the industry is promoted from the perspective of open patent data, a complex of technology, law and economy. Through the technology and economy behind the patent data to ensure its reliability and authenticity, this paper puts forward suggestions and prospects for the practitioners in the field of nuclear facilities decommissioning from the aspects of how to combine the research and analysis of patent situation with the development of the industry.
Aerosol Removal by a Heat Exchanger of Passive Containment Cooling System
Authors: Liang Yangyang , Lu Junjing , Zhang Tianqi , Han Xu , Yuan Yidan
Published: October 2021

The aerosol removal mechanism and capacity related to passive containment cooling system (PCS) for PWR reactors during the severe accident is an important and interesting topic among researchers, which can enhance the natural removal of suspended aerosols, reducing the radioactivity that may leak into the environment. In this study, the heat transfer and aerosol deposition models of a single heat tube of PCS heat exchanger were established. The four natural removal mechanisms (gravity deposition, Brownian diffusion, diffusiophoresis, and thermophoresis) on the tube surface were focused on. The results show that the diffusiophoresis and thermophoresis deposition velocities decease along the flow direction of cooling water inside the heat tube. The maximum total removal coefficient locates at the position that directly above the tube inlet. Aerosol particles with large diameter cannot deposit on the bottom of the heat tube under the certain thermal-hydraulic condition. With the continuing operation of PCS, the total removal coefficient for a PCS heat exchanger decreases rapidly at the beginning and then decreases slowly, while the diffusiophoresis effect is always the dominant for the natural removal.
A Study of Heat Transfer and Flow Characteristics Under Non-Uniform Thermal Boundary Condition
Authors: Wang Qiang , Xu Yuting , Wang He
Published: October 2021

In the industry, most of the heating equipment of the thermal system could not achieve complete constant heating in theory. It is more complicated in nuclear power system. Even if the effect of burn up life is not considered, the distribution of power still presents a special function distribution. However, in order to simplify the problem, the heat release unit is often replaced by uniform heat flux. In the process of system characteristics research, the effect of local heat transfer usually will not have a great impact on the overall performance of the device. But from the perspective of safety design, once the local heat pipe is burned out, it will directly restrict and determine the design limit of the whole device. Therefore, it is of great engineering significance to study the characteristics of unconventional thermal boundary conditions.

In order to reveal the principle of heat transfer character under specific boundary conditions, a series of investigations are conducted. Based on the conventional test section with uniform heat flux, steady-state and transient conduction differential equations are derived theoretically and extended to the experimental channel with axial cosine heating. The theoretical deduction results indicate that the amplitude and phase difference of wall temperature are influenced by Fourier number when flowrate oscillates. Meanwhile, numerical simulation is conducted on heating tube with different thermal boundary conditions. The wall temperature field of axial nonuniform heating channel is studied under steady and unsteady flow. It includes the position of peak wall temperature, excess temperature, and the non-linear distribution of fluid temperature along the channel. The results indicate that a great difference exist between uniform and non-uniform heating flux. The simulation of heat and flow characteristics in reactor core is not suggested to use conventional test channel with uniform heat flux. The related investigations could lend some experience to nuclear safety design.
The Safety Analysis of the Design of the Reactor Coolant Pump Heat Shield in Qinshan Nuclear Power Plant
Authors: Wang Zilong , Liu Deyi , Zhao Ming , Wang Li , Xu Jie
Published: October 2021

The design of the reactor coolant pump heat shield of Qinshan Nuclear Power Plant in China National Nuclear Power Co., Ltd. is a weakness in the design of the EDF standard 900MW nuclear power plant. When the reactor coolant pump heat shield is broken, and the single failure of the equipment cooling water system valve (RRI219VN (220VN) or RRI225VN (226VN) valve) is considered, it is inevitable that the high-pressure coolant will damage the equipment cooling system (RRI), resulting in a LOCA accident of the containment bypass type. The Unit 3 and Unit 4 of Qinshan Nuclear Power Plant were improved during the construction stage, and the upstream of the heat shield was changed to 2 check valves and the downstream was 2 electric valves. The modification hasn’t been performed in Unit 1 and Unit 2 of Qinshan Nuclear Power Plant. Qinshan Nuclear Power Plant has entrusted the research institute to carry out the plan design, and the on-site verification of the plan was limited by the on-site space and could not be implemented. There are also factors such as a large amount of work to be carried out by the main control room board replacement, and insufficient time for outage implementation. This article makes a safety analysis of this technical modification using PSA method to determine the necessity of the technical modification.
Eigenvalue Analysis of Well-Posedness of Two-Fluid Single Pressure Model With Virtual Mass Force and Interfacial Pressure
Authors: Chao Fei , Yang Wen , Li Longze , Qiu Jinrong , Shan Jianqiang
Published: October 2021

It is well know that the two-fluid single pressure model is currently widely used to analyze reactor transient accidents such as LOCA. Current mainstream reactor safety analysis codes such as RALAP5 and CATHARE are based on this single pressure model. However this model has been proved to be ill posed in the sense that the equation system is non-hyperbolic which will lead to numerical unphysical oscillation. Currently reactor safety analysis codes use the interfacial pressure and virtual mass force to solve the ill-posed non-hyperbolicity issue. The development of China’s self-owned reactor high fidelity system analysis code has attracted much attention in recent years, and studying on the ill-posedness of the two-fluid model and the improvement of the ill-posedness is an important basis to analyze these reactor accidents. In this paper, the ill-posedness regions of the two-fluid single pressure model is first investigated using the eigenvalue analysis method based on the Cauchy problem with initial conditions. Then the effect of the virtual mass force and the interfacial pressure is studied by this eigenvalue analysis method, and three types of virtual mass term in the momentum equation are discussed. The results show that the appropriate virtual mass force and interfacial pressure can well improve the ill-posedness of the single pressure model, and the appropriate combination of them can significantly improve the ill-posedness.
Modeling and Sensitivity Analysis of the Sodium-Water Reaction Accident in Parallel Channels
Authors: Luo Gang , Sun Peiwei , Bai Xi , Cao Huasong , Wang Kai , Zhu Huanjun
Published: October 2021

A pressure wave propagation across the second loop of a sodium-cooled fast reactor may lead to severe damage to the pipes and the equipment due to a large leakage sodium-water reaction accident. Therefore, the pressure source and the pressure wave propagation calculation and analysis can be significant for a sodium-cooled fast reactor’s design and operation.

A mathematical model with code was built to calculate and predict both the pressure source and the pressure wave propagation after the large leak sodium-water reaction accident occurred in an SFR steam generator.

In the pressure source model, the sodium-water reaction produces hydrogen and squeezes sodium nearby. It is assumed that the sodium-water reaction carries out instantaneously, and a spherical-to-column bubble is assumed to grow up at the reaction zone. Under such assumptions, the pressure source model can obtain the hydrogen bubble’s volume and pressure in the steam generator.

In the pressure wave propagation model, flow is assumed to be one-dimensional. With the basic theory of mass conservation, momentum conservation, and the Isentropic relation of the working fluid, the numerical solution is calculated by a grid divided by the method of characteristics, considering both precision and efficiency.

However, plenty of parameters, both of the pressure source and the parallel channels, can affect the accident in different ways and scales. Considering that to practice a large leakage sodium-water reaction experiment can be somewhat difficult because of its danger, inconvenience, and high cost, it is quite essential to conduct a sensitivity analysis through the program for the purpose of facilitating the design of an SFR.

Since most sodium-cooled fast reactor designs are multimodules, a parallel channel structure with multiple pressure boundaries is modeled for sensitivity analysis. Sensitivity analysis was performed on multiple sets of important parameters, including different hydrogen bubble properties, different pressure boundaries, and different equipment types. Following the sensitivity study, the code can be used for the accident analysis in a future commercial demonstration fast reactor by indicating various parameter guidance for the design and construction.
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

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

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

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

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

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

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

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

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

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

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

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

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

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.