bluuebunny/crossref_metadata_2025_split huggingface dataset:
Modeling Axial Relocation of Fragmented Fuel During Loss of Coolant Conditions by Using ABAQUS
Authors: Ma Zehua , Shirvan Koroush , Li Wei , Wu Yingwei 
Published: October 2020

Abstract
In a light-water reactor, during normal operating condition, the UO2 nuclear fuel pellets undergo fragmentation primarily due to presence of thermal stresses, fission gas development and pellet-clad mechanical interaction. Under Loss of Coolant Accident (LOCA) conditions, a portion of fuel fragments can freely move downwards to the ballooning region due to the significant cladding deformation. The fuel relocation can localize the heat load and in turn accelerate the cladding balloon and burst process. Cladding burst is of great concern because of the potential for fuel dispersal into coolant and clad structural stability. In our work, we built up a finite element model considering cladding balloon, fuel relocation and its resultant thermal feedback during LOCA condition with ABAQUS. The clad balloon model includes phase transformation, swelling, thermal and irradiation creep, irradiation hardening and annealing and other important thermal-mechanical properties. The mass of relocation model was verified against the analytical cases of single balloon and twin balloons. The cladding balloon model combined with fuel thermal conductivity degradation was verified against fuel performance code, FRAPTRAN. Finally, with the evolution of pellet-cladding gap, the fuel mass relocation was calculated and compared against the IFA-650.4 transient test from the Halden reactor.
Experimental Study on Two Different Gas-Liquid Separators Under Different Flow Patterns
Authors: Zeng Xiaobo , Yan Changqi , Fan Guangming , Cheng Jie , Xu Junxiu , Liu Antai 
Published: October 2020

Abstract
Gas-liquid separation is widely used in many fields, such as nuclear energy and petroleum resources. And the gas-liquid mixture separated gradually shows the characteristic of wide range of gas void fraction and variable flow patterns. However, the current separators only suit for narrow range of gas void fraction or single flow patterns. In this research, two different new type separators using centrifugal technology were designed and an experimental system was constructed to test the two separators using dry air and water under different flow patterns, including bubble, slug and churn flow. One was called inline separator consisting of three swirls and another was called double-layer cylinder separator composed of a central tube, a swirl and an outer tube. The results show that the separation performance of the inline separator was sensitive to flow patterns and the two-layer cylinder separator keeps high efficiency in different flow patterns. In bubble flow and slug flow patterns, the two separators kept high efficiency, while the oscillation of the gas core in the inline separator aggravated under slug flow condition. When increasing the gas void fraction, the turbulence of the churn flow led to the diameter of the gas core change drastically and reduce separation efficiency significantly.
Development of Approaches to Simulate Fuel Rod Destruction With Different Fuel Type
Authors: Usov Eduard , Pribaturin Nikolay , Chukhno Vladimir , Klimonov Ilya , Butov Anton , Kudashov Ivan , Gafiyatullin Albert , Lobanov Pavel 
Published: October 2020

Abstract
Due to the revival of interest to the development of fast reactors cooled by liquid metals, the problem of carrying out theoretical research in support of their safety is actual. A detailed calculation of all stages of the accident from the beginning to the end requires knowledge of the laws for modeling physical processes occurring in the reactor in an emergency. The most serious are accidents with the destruction of the core. Simulation of severe accident in nuclear reactor is the key element in safety analysis of nuclear power plants. Destruction of fuel rods is one of the most important processes that should be calculated during core degradation. For different type of fuels the mechanism of the degradation are different too. For example, oxide and metallic fuels usually melt congruently at high temperature, but nitride fuel dissociates. The main objective of the proposed research is developing of models and numerical algorithms for calculation fuel rods destruction with oxide, metallic and nitride fuels. The models of the destruction processes and some calculation results are presented in the paper. The processes are investigated for the first phase of severe accidents covering the period from the onset of fuel-rod melting to the melt escape from the core center.
Numerical Study on Discharging Characteristics of Pebble Cluster Flow in Pebble Bed
Authors: Ge Liang , Gui Nan , Yang Xingtuan , Tu Jiyuan , Jiang Shengyao 
Published: October 2020

Abstract
To better understand the flow features of pebble cluster in pebble bed, discharging of the pebble cluster were simulated by DEM. The pebble entangled cluster was composed of eight particles connected by rigid bonds and the simulated cluster models are divided into two types: axisymmetric u-particle and distorted z-particle. The simulation starts with the closed discharge outlet and the bonded clusters with different ID are randomly added from the entrance section. The pebbles fall freely and accumulate freely in the pebble bed. The discharge hole opens after all the pebbles being stationary for a period. Then the pebbles are discharged from the pebble bed under gravity. The discharging process is time-dependent bulk-movement behavior. There is not much mixing between layers on the boundary.
The vertical end makes the packing loose, but also intensifies the interaction between particles due to entanglement. Consequently, the discharge features of pebble clusters of different included angles were quantified. The results show that the pebble discharging speeds depend on entanglement angle (α of u-particle and η of z-particle) and discharging outlet diameter. A large included angle may play the role of retarding or inhibiting the discharging flowrate. Therefore, the entanglement of particles component also always plays the key role of retarding the discharge.
Weakly Supervised Deep Neural Network for Bearing Fault Diagnosis
Authors: Miki Daisuke , Demachi Kazuyuki 
Published: October 2020

Abstract
Bearings are one of the main components of rotating machinery, and their failure is one of the most common cause of mechanical failure. Therefore, many fault detection methods based on artificial intelligence, such as machine learning and deep learning, have been proposed. Particularly, with recent advances in deep learning, many anomaly detection methods based on deep neural networks (DNN) have been proposed. DNNs provide high-performance recognition and are easy to implement; however, optimizing DNNs require large annotated datasets. Additionally, the annotation of time-series data, such as abnormal vibration signals, is time consuming. To solve these problems, we proposed a method to automatically extract features from abnormal vibration signals from the time-series data. In this research, we propose a new DNN training method and fault detection method inspired by multi-instance learning. Additionally, we propose a new loss function for optimizing the DNN model that identifies anomalies from a time-series data. Furthermore, to evaluate the feasibility of automatic feature extraction from vibration signal data using the proposed method, we conducted experiments to determine whether anomalies could be detected, identified, and localized in published datasets.
Benchmarking an AI-Guided Reasoning-Based Operator Support System on the Three Mile Island Accident Scenario
Authors: Hanna Botros N. , Son Tran C. , Dinh Nam T. 
Published: October 2020

Abstract
In the Nuclear Power Plant (NPP) control room, the operators’ performance in emergencies is impacted by the need to monitor many indicators on the control room boards, the limited time to interact with dynamic events, and the incompleteness of the operator’s knowledge. Recent research has been directed toward increasing the level of automation in the NPP system by employing modern AI techniques that support the operator’s decisions. In previous work, the authors have employed a novel AI-guided declarative approach (namely, Answer Set Programming (ASP)) to represent and reason with human qualitative knowledge. This represented knowledge is structured to form a reasoning-based operator support system that assists the operator and compensates for any knowledge incompleteness by performing reasoning to diagnose failures and recommend executing actions in real time. A general ASP code structure has been proposed and tested against simple scenarios, e.g., diagnosis of pump failures that result in loss of flow transients and generating the needed plans for resolving the issue of stuck valves in the secondary loop.
In this work, we investigate the potential of the previously proposed ASP structure by applying ASP to a realistic case study of the Three Mile Island, Unit 2 (TMI-2) accident event sequence (in particular, the first 142 minutes). The TMI scenario presents many challenges for a reasoning system, including a large number of variables, the complexity of the scenario, and the misleading readings. The capability of the ASP-based reasoning system is tested for diagnosis and recommending actions throughout the scenario. This paper is the first work to test and demonstrate the capability of an automated reasoning system by applying it to a realistic nuclear accident scenario, such as the TMI-2 accident.
Metamorphic Testing on Nuclide Inventory Tool
Authors: Li Meng , Yan Shiyu , Yang Xiaohua , Liu Jie 
Published: October 2020

Abstract
The Chebyshev rational approximation method (CRAM) is an important numerical solution algorithm for the burnup equation. Since high complexity of nuclide depletion calculation, its outputs are almost impossible to predict accurately. Therefore, the traditional testing methods are inapplicable, even invalid. The metamorphic testing (MT) is one of promising method to solve such typical testing oracle problem. However, the absent meta-morphic relations (MRs) seriously hinders its application.
In response to the nuclear software development process, we established a nuclear MR hierarchical model (NMHM). From top to bottom, it divides MRs into three layers: physics, algorithm and code. From the burnup equation and the CRAM, we carried out a group of MRs according to the NMHM. Employing these MRs, we verified the Nuclide Inventory Tool (NUIT) which implements the CRAM by MT. The MRs adopted can certainly be used for other CRAM programs, and the NMHM will extend to more nuclear science software for guiding the MR identification and classification.
Analysis of the Behavior of Droplet Impinging on a Curved Dry Wall in a Rotating Flow Field
Authors: Ouyang Shuo , Xiong Zhenqin 
Published: October 2020

Abstract
Steam water separator is the core equipment of nuclear power plant. It is very vital for improving the efficiency of the steam separator to study the impact characteristics between the droplets and the curved dry wall of the steam separator under the action of the rotating air flow. In this paper, the characteristics of droplet impinging on the dry wall in the rotating flow field are analyzed by establishing a visualization experimental device. A high-speed camera was used to capture the impact of droplets with different diameters on the dry wall of a non-wetting curved surface at different gas velocities. At the same time, using image processing tool in MATLAB to obtain image boundary information. The characteristics of spreading factor, droplet deformation factor and initial diffusion velocity of droplets impacting the surface dry wall under different wind speeds are studied.
A Preliminary Study on Neutronic Performance of the Spallation Target With the Proton Beam Variation
Authors: Zhou Junjie , Zeng Qin , Yang Jinchen , Yang Yi , Shi Ying , Jiang Yanyi 
Published: October 2020

Abstract
Spallation particles can be utilized in different fields such as neutron scattering studies, external source for burning spent fuel as well as running subcritical reactors. Accelerator driven subcritical system (ADS) has been considered as an advanced nuclear waste transmutation facility with inherent safety feature. In this paper, the neutronic performance of the spallation target with the proton beam variation was investigated. Monte Carlo code FLUKA has been used to calculate neutronic parameters including: spallation neutron flux and energy deposition. As the beam position is shifted towards the positive direction of the Y-axis, the neutron flux and energy deposition increased significantly along the positive direction of the Y-axis, while the flux and energy deposition in the opposite direction decreased. Such a change may lead to a sharp increase of the neutron flux of fuel components near the beam target in subcritical reactor. But as the beam shifts further and further, the neutron flux in the positive direction declines compared to the previous data. The calculation and analyzing the data obtained from proton beam variation shows that it has a great influence on the physical properties of the neutron spallation target.
Guided Waves in Pitch-Catch Configuration: A Theoretical Framework to Steer Toward Experimental Tests
Authors: Cordella Francesco , Cappelli Mauro , Bertoncini Francesco 
Published: October 2020

Abstract
Guided waves testing allows a long-range screening in pipes of different types and represents an effective and powerful non-destructing technique for defect detections using a limited number of points of measures. This kind of testing hence represents an appealing technique not only for the Oil and Gas industries but also for the Nuclear Industry, in particular regarding the Structural Health Monitoring of Nuclear Power Plants components. Another point of strength of this technique is that it can be applied in different configurations as the pulse-echo (the same probe is used both for transmission and signal receiving) or the pitch-catch (two symmetric probes are used one for the signal transmission and the second one for the signal receiving). In this way, the guided wave testing with magnetostrictive sensors can be reliably used for the short and long-term monitoring of Nuclear Power Plants components.
The objective of this paper is to establish a strong theoretical background to pave the way for a robust experimental investigation. In particular, after the characterization through a general theoretical analysis, the focus is on a real steam discharge pipe with a high mechanical complexity used for many years in a research facility and now dismissed. The experimental method applied is the pitch-catch configuration of two magnetostrictive sensors. Preliminary experimental results conducted on a real complex steam discharge pipe are consistent with the theoretical analysis.
Numerical Investigation on Heat Transfer to Supercritical Water Flowing Upward in a 4-M Long Bare Vertical Circular Tube
Authors: Yang Dong , Wu Qixian , Chen Lin , Pioro Igor 
Published: October 2020

Abstract
Thermal efficiency and safety of Generation-IV nuclear-power-reactor concept - Supercritical Water-cooled Reactor (SCWR) depend on solid knowledge of specifics of SCW thermophysical properties and heat transfer within these conditions. As a preliminary, but conservative approach to uncover these specifics is analysis of experimental data obtained in bare tubes including numerical investigation. This paper presents the numerical investigation, based on computational fluid dynamics, of the heat-transfer characteristics of SCW flow in a 4-m long circular tube (ID = 10 mm). The flow and heat-transfer mechanism of SCW in the vertical tube under the influence of buoyancy and flow acceleration are analyzed. Results of numerical simulation predict the experimental data with reasonable accuracy. The results indicated that in the region of q/G &gt; 0.4 kJ/kg, the wall temperature distribution tends to be non-linear, and heat transfer may deteriorate. When Tb &lt; Tpc &lt; Tw, internal wall temperature shows peaks, which corresponds to heat-transfer deterioration. Meanwhile the position, where the deterioration occurs is continuously moved forward to the inlet as the heat flux increases. Velocity changes near the wall show an M shape according to mass conservation for the density change.
Optimization of In-Core Detector Locations in AHWR Using Bond Energy Algorithm
Authors: Balajiranganathan Anupreethi , Gupta Anurag , Kannan Umasankari , Tiwari Akhilanand Pati 
Published: October 2020

Abstract
A solution to optimization of in-core detectors placement for Advanced Heavy Water Reactor (AHWR) has been attempted. AHWR houses in-core detector units with Self-Powered Neutron Detectors (SPND) distributed axially and their measurement serves as an input to Online Flux Mapping System (OFMS) to monitor the three-dimensional neutron flux distribution. There is a requirement of placing these in-core detectors at optimum locations to retrieve maximum information about the reactor while keeping their number to the minimum. This paper attempts to optimize SPND placement through the application of Bond Energy Algorithm (BEA), a clustering technique which groups the SPNDs based on correlation. This works on the concept of grouping strongly correlated SPNDs into blocks and choosing one SPND from each block as the optimal location. The higher the uncorrelation among optimal SPNDs, the higher the independent information retrieved about the actual configuration of the reactor. It can be inferred from this work that the number and location of SPNDs are highly dependent on the initial set of SPND locations and the correlation threshold. It can be seen that as the correlation threshold increases, the number of optimal locations increases. The obtained optimal locations have been validated for various operational reactor configurations using different Flux Mapping Algorithms (FMA).
Thermal Safety Margin Calculation of the MP-2 Experiment in the Advanced Test Reactor
Authors: Hawkes Grant L. 
Published: October 2020

Abstract
The Mini-Plate 2 (MP-2) irradiation test is a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The experiment is a drop-in test where small aluminum-clad fuel plate samples (mini plates) are cooled directly by the ATR Primary Coolant System (PCS) water. The MP-2 fuel plate experiment will be irradiated in several different irradiation locations of the ATR. This fueled experiment contains aluminum-clad fuel mini plates consisting of monolithic U-Mo. Four different types of fuel plates with fuel meat thickness and cladding are part of the MP-2 test. A thermal analysis has been performed on the MP-2 experiment. A method for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) during a reactivity transient using the commercial finite element and heat transfer code ABAQUS is discussed. At the start of an ATR cycle the heat generation rate of the fueled experiment is high and the heat rate multiplier from the outer shim control cylinders is low, while the reverse is true at the end of the ATR cycle. Thermal analyses at 10-day increments during the cycle calculate the DNBR and FIR during a reactivity transient. This technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node at various times during the ATR cycle. Heat rates vary with time during the transient calculations that are provided by a detailed physics analysis. Oxide growth on the fuel plates is also incorporated.
Results from the transient calculations are displayed with the ABAQUS post processor. By calculating these parameters at each location in the finite element model, conservatism is replaced with accuracy. This allows for a greater margin for the thermal hydraulic safety parameters.
CFD Preliminary Assessment of the ALFRED FA Thermal-Hydraulics
Authors: Marinari R. , Di Piazza I. , Tarantino M. , Grasso G. , Frignani M. 
Published: October 2020

Abstract
In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the coolability of the Fuel Assembly in nominal condition is of central interest.
The Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED) is a 300 MWth pool-type reactor aimed at demonstrating the safe deployment of the Generation IV LFR technology. The ALFRED design, currently being developed by the Fostering ALFRED Construction international consortium, is based on prototypical solutions intended to be used in the next generation of lead-cooled Small Modular Reactors.
Within the scope of FALCON and in the frame of investigating the thermal-hydraulics of the ALFRED core, a CFD computational model of the general Fuel Assembly (FA) is built looking for the assessment of its thermal field in nominal flow conditions both for the average FA and the hottest one. Starting from the experience in this kind of simulations and in experimental work, the whole model of the ALFRED Fuel Assembly is first presented and calculation of flow and temperature field in nominal conditions is carried out. Results showed that the thermal hydraulic field predicted in the average FA by the code is in good agreement with analytical correlations and the temperature field on the pin clad is acceptable for clad material temperature constraint. About the results on the hot FA test case, the CFD results highlighted a peak temperature on the clad close to the clad temperature constraint. This result led to an upgrade of the mass flow distribution among the FA for achieving a 20% mass flow increase in the hottest one that guarantees higher temperature margin on the clad.
Calculation of Critical Heat Flux Using an Inverse Heat Transfer Method to Support TREAT Experiment Analysis
Authors: Armstrong Robert , Folsom Charles , Hill Connie , Jensen Colby 
Published: October 2020

Abstract
Heat transfer between cladding and coolant during transient scenarios remains a critical area of uncertainty in understanding nuclear reactor safety. To advance the understanding of transient and accident scenarios involving critical heat flux (CHF), an in-pile experiment for the Transient Reactor Test facility (TREAT) at Idaho National Laboratory (INL) was developed. The experiment, named CHF-Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA), consists of a hollow borated stainless-steel heater rod submerged in a static water pool heated via the (n, α) reaction in boron-10. This paper presents a novel inverse heat transfer method to determine CHF by using the optimization and uncertainty software Dakota to calibrate a RELAP5-3D model of CHF-SERTTA to temperature measurements obtained from a thermocouple welded to the surface of the rod.