bluuebunny/crossref_metadata_2025 huggingface dataset:
Study on Measures During Loss of Normal Feedwater Accident for AP1000 NPP in Lower Power Operation
Authors: Ma Baisong , Miao Zhengqiang , Zhuang Yaping , Hou Weili
Published: October 2020

If an accident occurred in a plant in lower power operation, a series of problems could be caused by low residual heat and excessive removal capability of heat sinks. Taking a real accident as example, this paper discussed how to deal with loss of normal feedwater in AP1000 NPP during lower power operation. The level of Steam Generators (SG) will be rapidly reduced in case of Main Feedwater loss, and Startup Feedwater pump will start. If the reactor is not shut down immediately, the Startup Feedwater will not be enough to prevent SG level from falling, and automatic reactor trip will be inevitable. Therefore, in event of loss of Main Feedwater which cannot be recovered instantly, the reactor should be shut down manually to maintain SG level as high as possible. Appropriate measures should be taken to avoid excessive cooling after trip. Attention should be paid to SG level and feedwater flow to avoid Passive Residual Heat Removal Heat Exchanger (PRHR HX) actuation during regulating Startup Feedwater. Otherwise, PRHR HX may be actuated unexpectedly. With Reactor Coolant Pumps operating, PRHR HX heat removal capacity is strong. The Cold Leg temperature will drop rapidly because of PRHR HX and other heat sinks. In this case, the most effective measure to stabilize temperature of Cold legs is to isolate Streamline. Otherwise, structure integrity of the Reactor Pressure Vessel (RPV) will be challenged under Pressurized Thermal Shock (PTS) due to rapid drop of Cold leg temperature. In order to alleviate PTS, depressurization shall be performed and further cooling shall be suspended. Since PRHR HX is only connected to Loop 1, coolant in Hot Leg of Loop 2 is still kept at high temperature. Therefore, necessary measures shall be taken to prevent coolant flash in Loop 2 during the process of depressurization, which may fill the Pressurizer up with water. The RPV Head Vent valves perform safety-related function by preventing Pressurizer overfill in certain design basis events in AP1000 NPP. However, emergency letdown by opening the valves cannot effectively reduce Pressurizer level when coolant flashed in Loop 2. Because of mitigating PTS, opening RPV Head Vent valves, and steam condensation in Loop 2, the subcooling of the coolant at the core outlet is likely to drop below 6 °C or even 0 °C.
Study on the Effect of Different Factors of Displacement Cascades in Alpha-Fe by Molecular Dynamics Simulations
Authors: Lin Pandong , Nie Junfeng , Liu Meidan
Published: October 2020

As the key component of RPV steel, α-Fe is under neutron irradiation during its long-term service, and lattice atoms of α-Fe are knocked by neutrons, which leads to irradiation damage. In this paper, molecular dynamics method is conducted to investigate the effect of temperature, vacancy concentration and tensile strain on irradiation-induced damage by displacement cascade simulations in α-Fe. The simulations are performed with primary knock-on atom energies ranging from 0.1 to 5 keV, temperature ranging from 100 to 500K, vacancy concentration ranging from 0% to 1% and applied tensile strain ranging from 0 to 5%. The time evolution of defects produced during displacement cascade can be obtained where the surviving number of Frenkel pairs increases rapidly at first, then decrease and comes to stability finally. The influence of these factors on defect production can be concluded as following: The increase of PKA energy, vacancy concentration and applied tensile strain can lead to the increase of defect numbers. In contrast, the increase of temperature decreases the defect numbers. Vacancies and interstitials cluster size distributions are varied in different case. The results are meaningful to describe some microcosmic mechanisms of RPV steels in nuclear system.
Multi-Group Staggered Tests for Highly Redundant Systems
Authors: Soga Shota
Published: October 2020

A reactor protection system (RPS) in a boiling water reactor (BWR) is a unique system since it represents high redundancy compared with other safety-related systems in the BWR. Owning to its high redundancy, NUREG-5500 vol.3 showed that unreliability of the General Electric BWR4-type RPS is dominated by common-cause failures (CCFs) in its highly redundant components. Since the staggered test scheme can reduce the period of latent CCF states of redundant standby components, the scheme is a good candidate to improve the reliability of the RPS. However, the scheme is impracticable for the RPS because of its high redundancy.

In this paper, a new concept of a “multi-group staggered test scheme” for a highly redundant system is proposed. In the multi-group staggered test scheme, components are arranged into several groups, and all components in a group are tested simultaneously. This paper presents a preliminary example of the new scheme that is applied to control rod notch tests, in which all withdrawn control rods are moved at least one notch. It is shown that the new scheme can reduce the CCF risk in RPS.
Experimental Investigation of Particle Decontamination Efficiency in a Single-Bubble by Pool Scrubbing
Authors: Fujiwara Kota , Nakamura Yuki , Yoshida Kohei , Kaneko Akiko , Abe Yutaka
Published: October 2020

Nuclear power plant (NPP) safety has become a public issue since the Fukushima daiichi NPP accident. In order to evaluate the risks caused by severe accidents (SAs), it is very important to understand the on-site source term events. One of the important unsolved source term events is the decontamination efficiency of fission products (FPs) in the suppression chamber by pool scrubbing. Therefore, a mechanistic model to analyze the particle decontamination efficiency by pool scrubbing is highly regarded. Despite the demand, particle decontamination mechanism by pool scrubbing has never been understood due to the complexity of phenomena.

In our experiment, we aim to develop a reliable mechanistic model to evaluate particle decontamination efficiency of pool scrubbing by conducting separate effect tests. As to obtain the fundamental process of particle decontamination from gas to liquid-phase, we focused on decontamination factor (DF) of particle from a single bubble. However, it is very difficult to calculate the initial particle concentration inside the bubble. Therefore, in our experiment, we developed a method to measure the internal particle concentration inside the bubble by combining image processing and particle measurement.

By using the experimental results, we succeeded to obtain reasonable DF for glycerin particles and CsI particles as a simulant particle for FPs. From the experimental results, detailed particle decontamination efficiency for various submergence were measured. The results tend show that DF increase linearly as submergence increases which suggests that DF is constant on bubble rise region. Moreover, the fact that glycerin particle with larger particle diameter takes a higher value shows that particle diameter significantly affects DF.
Review of Application on Dynamic Fault Tree Method in Nuclear Power Plants
Authors: Guohua Wu , Diping Yuan , Yiqing Xiao , Jiaxin Wang
Published: October 2020

Fault tree analysis (FTA) is one of the most important methods of probabilistic risk assessment (PRA). The fault state of the system is taken. While traditional FTA is based on static failure model. FTA is not applicable for systems that include redundant, sequence-related systems. At the same time, nuclear power plants (NPPs) contains a large number of redundant equipment, and FTA is difficult to solve these dynamic problems. Therefore, it is necessary to use dynamic fault tree analysis (DFTA) for PRA. In DFTA research, the modular analysis method was first proposed. The modular method divides the dynamic fault tree into a dynamic fault tree and a static fault tree. Among them, the dynamic fault tree is analyzed using a Markov chain, and the static fault tree is studied using a binary decision diagrams method. However, the shortcomings are that when the system is complicated, the information explosion in the Markov chain is appeared. To solve this problem, a dynamic fault tree is transformed into a Bayesian network. At the same time, to verify the feasibility of the method, Monte Carlo random sampling was used to evaluate the method. Other methods are relatively infrequently studied.

In this paper, firstly, status of dynamic fault trees has been investigated. Secondly, each method is analyzed and the problems of dynamic fault tree are described. Finally, a survey and analysis on the dynamic fault tree is finished, and the main problems of the dynamic fault tree are: information explosion; the lack of commercial software to apply to engineering. Through this review, we hope to play a certain guiding role in the subsequent research on dynamic fault trees.
Development of a Flow Network Calculation Code (FNCC) for High Temperature Gas-Cooled Reactors (HTGRs)
Authors: Aoki Takeshi , Isaka Kazuyoshi , Sato Hiroyuki , Ohashi Hirofumi
Published: October 2020

A core of the prismatic high temperature gas-cooled reactor (HTGR) consists of fuel blocks, control rod guide blocks and replaceable reflector blocks with surrounding permanent reflector blocks. The core flowrate available for effective cooling of fuels, fuel blocks and control rod guide blocks has to be evaluated to confirm that maximum fuel temperature satisfies fuel design limits. In the evaluation, unintended flows such as gap flows between columns, cross flows between column layers and gap flows between permanent reflectors should be analyzed for the prismatic-type HTGR design. The flow distribution has to be evaluated preciously with minimizing the unintended flows.

In the HTGR design, flow distribution analysis has to take into account the interaction of the gaps between the graphite structures forming unintended flow paths and flow conditions such as coolant temperature and pressure by thermal and radiation deformations of the graphite structure.

In the present study, a user-friendly flow network calculation code (FNCC) has been developed on the basis of experiences of High Temperature engineering Test Reactor (HTTR) design for HTGR design with enhanced compatibility with other HTGR design codes and with considering graphite block deformation in iteration process without manual control. The code is able to solve the flow distribution based on conservation equations on the mass, momentum and energy by converging iteration calculations along with update of gap information in each calculation step. The validation result using FNCC provides good agreement with a result of the one-column flow distribution experiment.
Synergetic Oxidation in Alkaline In-Situ Leaching Uranium: A Preliminary Case Study
Authors: Liao Wensheng , Que Weimin , Wang Liming , Du Zhiming
Published: October 2020

In alkaline in-situ leaching uranium, oxygen is the most common oxidizer with bicarbonate as a complexing agent. For those sandstone uranium deposits with strongly reductive capacity or complicated hydrogeological environment, the oxidation by oxygen is low efficiency. An efficient leaching method, therefore, is needed for these uranium deposits. In this study, a typical sandstone uranium deposit which characterizes with high TDS and high chloride content in groundwater and intractable uranium leach is selected to investigate the effects of synergetic oxidation by a strong oxidant with oxygen. Based on the research on batch leach, pressure leach and field trials, the oxidants such as hydrogen peroxide, potassium permanganate and sodium dichloroisocyanurate (NaDCC) are tested. The results of pressure batch leach indicate that synergetic oxidization is achieved by NaDCC in oxygen leaching process. Leaching tests indicate that a minor oxidizer of NaDCC shows good synergetic oxidization with oxygen and leaching effects on uranium minerals. The results also demonstrate that hydrogen peroxide shows no oxidation effects when it is used as a single oxidant. While potassium permanganate shows good oxidation on uranium when it is used as a single oxidant, however, it leads inhibiting effects on oxygen oxidation on uranium minerals. The further field tests are conducted to study the synergetic effects of oxygen with and without sodium dichloroisocyanurate. The preliminary results indicate that a fast leach is observed by the composite oxidants in early stage while no synergetic leach is found after 200 days. Further studies should be conducted in laboratory experiments and pilot scale tests for its potential applications.
Numerical Study of Collision Behavior of Melt Drops During Fuel-Coolant Interaction
Authors: Wen Panpan , Li Gen , Gao Jinchen , Li Yupeng , Yamaji Akifumi , Yan Junjie
Published: October 2020

The collision dynamics between two droplets plays an important role in various disciplines of nature and practical interests, such as fuel-coolant interaction (FCI), fuel combustion in engines, and various spraying process. FCI presents in nuclear reactor severe accident when the melt relocates into the coolant in the lower head with violent disturbance and vigorous heat transfer. The purpose of this study is to investigate the collision behavior of melt droplets during fuel-coolant interaction. The collision of two equal-sized droplets has been simulated in 3D by using the volume of fluid (VOF) and adaptive mesh refinement method. The numerical simulations of tetradecane droplet collision were carried out to validate the numerical methods. The results showed good agreement with the experiments. Furthermore, the simulations of uranium dioxide (UO2) droplets collision in coolant were carried out. The results showed that the contact area between droplets and coolant increased with time first and then decreased. With the increase of Weber number, the contact area of maximum in the droplet collision increased. Break happened in the later period and many child droplets formed. The number of child droplets increased with the increase of Weber number. In addition, the size distribution of little droplets was investigated.
Proportional-Integral Disturbance Observer of Nuclear Reactors
Authors: Dong Zhe
Published: October 2020

A proportional-integral disturbance observer (PI-DO) for monitoring nuclear reactors is newly proposed, which is driven by the measurements of neutron flux and coolant temperature at reactor inlet as well as their integrations. This PI-DO provides a globally asymptotic estimation with a bounded steady-state error for the reactor key process variables as well as the total disturbances in channels of the neutron kinetics and primary coolant thermal-hydraulics. Moreover, the PI-DO is applied to reconstruct the unmeasurable state variables and total disturbances of a nuclear heating reactor (NHR). Numerical simulation results not only verify the theoretic analysis but also show both the satisfactory performance and the influence of observer parameters.
Experimental Study on Flow Patterns of Decaying Swirling Gas-Liquid Flow in a Horizontal Pipe
Authors: Liu Shuai , Liu Li , Zhang Jiarong , Gu Hanyang
Published: October 2020

Swirling flow is one of the well-recognized techniques to control the working process. This special flow is widely adopted in swirl vane separators in nuclear steam generator (SG) for water droplet separation and the fission gas removal system in Thorium Molten Salt Reactor (TMSR) for gas bubble separation. Since the parameters such as separation efficiency, pressure drop and mass and heat transfer rate are strongly dependent on the flow pattern, the accurate prediction of flow patterns and their transitions is extremely important for the proper design, operation and optimization of swirling two-phase flow systems. In this paper, using air and water as working fluids, a visualization experiment is carried out to study the gas-liquid flow in a horizontal pipe containing a swirler with four helical vanes. The test pipe is 5 m in length and 30 mm in diameter. Firstly, five typical flow patterns of swirling gas-liquid flow at the outlet of the swirler are classified and defined, these being spiral chain, swirling gas column, swirling intermittent, swirling annular and swirling ribbon flow. Being affected by the different gas and liquid flow rate of non-swirling flow, it is found that the same non-swirling flow can change into different swirling flow patterns. After that, the evolution of various swirling flow patterns along the streamwise direction is analyzed considering the influence of swirl attenuation. The results indicate that the same swirling flow pattern can transform into a variety of swirling flow patterns and subsequent non-swirling flow patterns. Finally, the flow pattern maps at different positions downstream of the swirler are presented.
Friction Factor Correlation for Molten Metals Flow in a 3-Rod Bundle in Cylindrical Shroud, With Smooth and Scalloped Inner Wall
Authors: Schriener Timothy , El-Genk Mohamed S.
Published: October 2020

Thermal-hydraulic experiments to help develop and qualify new fuel designs for advanced liquid metal cooled reactors could use reduced-size triangular pitch, 3 rods within a circular shroud. In such an arrangement, scalloping the shroud wall helps achieve lateral flow uniformity and high velocity, but at the expense of increasing the pressure losses. Owing to the practical challenges of conducting experimental investigations with molten lead, Computational Fluid Dynamics (CFD) could a practical alternative to generate large enough database to develop a friction factor correlation. This work performed CFD analyses of molten lead flow in 3-rod bundle with circular shroud to investigate the effect of scalloping the shroud wall on the lateral flow uniformity and pressure. The developed friction factor correlation is very smooth, spanning the laminar, transition, and turbulent flow regions. Scalloping the shroud wall increases the pressure losses, but not the friction factor in the transition and turbulent flow regions. In the laminar flow region, however, scalloping the shroud wall only slightly increases the friction factor.
Emulation Methodology of Programmable Logic Controllers for Cybersecurity Applications
Authors: Fasano Raymond , Lamb Christopher , El Genk Mohamed , Schriener Timothy , Hahn Andrew
Published: October 2020

A programmable logic controller (PLC) emulation methodology can dramatically reduce the cost of high-fidelity operational technology (OT) network emulation without compromising specific functionality. A PLC emulation methodology is developed as part of an ongoing effort at the University of New Mexico’s Institute for Space and Nuclear Power Studies (UNM-ISNPS) in collaboration with Sandia National Laboratories (SNL) to develop an emulytic™ platform to support cybersecurity analyses of the instrumentation and control (I&C) systems of pressurized water reactors (PWRs). This methodology identifies and characterizes key physical and digital signatures of interest. The obtained and displayed digital signatures include the network response, traffic, and software version, while the selected physical signatures include the actuation response time and sampling time. An extensive validation analysis is performed to characterize the signatures of the real, hardware-based PLC and the emulated PLC. These signatures are then compared to quantify differences and identify optimum settings for the emulation fidelity.
Evaluation of Irradiation Embrittlement of the Weld Metals in the Chinese RPV
Authors: Cao Yupeng , He Yinbiao , Huang Yifeng , Wang Binxi , Yu Yan , Li Hui , Hui Hu
Published: October 2020

Reactor pressure vessel (RPV) operates under high temperatures and pressures and is exposed to relatively high neutron radiation. RPV is considered to be irreplaceable, which is the most limiting factor for the lifetime of a nuclear power plant. As the most severe ageing degradation mechanism in RPV materials, irradiation embrittlement is a major issue affecting the integrity through the service life of a RPV. Our previous paper (ASME PVP2019-93615[1]) introduced our project for assessment of irradiation embrittlement of the materials for the Chinese RPVs to verify the 60-year design life, in which the specimens made of the RPV base material manufactured in China, the SA-508 Gr.3 Cl.1 forging, and the different types of weld metals were irradiated in the high fluence engineering test reactor (HFETR). The paper analyzed extent irradiation damage of the forging in terms of mechanical properties. As another part of the project, this paper concentrates on the evaluation of the weld metals in the same project. Tensile tests, Charpy impact tests and fracture toughness tests by master curve approach were carried out for the three types of weld metals subjected to different irradiation fluences (2.6E19n/cm2, 8.9E19n/cm2). Comparison of the mechanical properties of the irradiated and the unirradiated materials is made. The irradiation resistance of the weld metals in our project is also compared with the data in the literatures.
Numerical Investigation of the Obstacle on Hydrogen Distribution in a Vessel
Authors: Wang Tianlin , Tong Lili , Cao Xuewu
Published: October 2020

Hydrogen combustion and explosion is an important safety issue in nuclear power plants (NPPs) containment during postulated severe accidents or in utilization of hydrogen. It is significant to understand the hydrogen flow and distribution in space for mitigating hydrogen risk. In this paper, a numerical model to investigate hydrogen flow and distribution in a vessel is established using computational fluid dynamics (CFD) code GASFLOW. Hydrogen is simulated by helium which is used to study the hydrogen distribution. The k-ε turbulent model is selected to establish the numerical model, and the numerical model which has no obstacle inside the vessel or includes the obstacle inside is verified under medium momentum conditions of injected gas by comparing numerical results with experimental data. Regardless of the presence of the obstacle in the vessel, helium stratification occurs under all momentum condition of injected gas. When the obstacle is present, it blocks the flow path of the injected helium to the upper space, then the helium volume concentration in upper space is lower than the condition that there is no obstacle in the vessel. As the initial Froude number increases from 0.19 to 19.29, the flow mechanism around the obstacle under high momentum condition of injected gas is different from that under medium or low momentum conditions. Consequently, the boundary of the helium stratification moves down, and the distribution of helium looks more uniform in most area of the vessel for high momentum conditions of injected gas.
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

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

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

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

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

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

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

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

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

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

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

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 > 0.4 kJ/kg, the wall temperature distribution tends to be non-linear, and heat transfer may deteriorate. When Tb < Tpc < 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

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

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

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

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.
Experimental Validation of CFD Models Capturing the Thermal-Hydraulics in Liquid Metal Cooled Reactor Plena
Authors: Ward Brendan , Hopkins Thomas , Bindra Hitesh
Published: October 2020

High fidelity velocity field experimental data in a liquid metal plenum is presented and compared with numerical simulations. While work has already been established for fluids like air and water, research on low Pr fluids (Pr ≪ 1) (e.g. liquid metals) has fewer experimental data sets with validation-quality data. Work in advanced reactors using liquid metal coolant requires validated numerical simulations for safety analyses. The Gallium Thermal-hydraulic Experiment (GaTE) facility is outfitted with acoustic backscattering measurement techniques to generate the high fidelity distributed flow field data in a liquid metal plenum (a 1/20th scale of the Department of Energy’s sodium cooled Advanced Burner Test Reactor design). The high spatial and temporal resolution of the sensors are required to capture the fluctuations of velocity to allow a more direct comparison to the numerical simulations. For these simulations the coupled mass and momentum equations under the large eddy simulation (LES) framework were solved with the wall-adapting local eddy-viscosity (WALE) model for sub-grid scale formulations. Since the temperature transients of interest for reactor safety have a period of about a minute in the GaTE system, there may not be enough time to allow statistical tools to check one-to-one correspondence. So the data collection period for both data sets was extended to allow convergence of the mean and a larger sample size for other statistics during system steady-state, isothermal tests. Two characteristic velocities of the plenum inlet barrel were investigated (U = 40, 60 mm/s; Re = 7,000, 11,000). Probability distributions show good agreement between experiment and simulation with the difference only in the low-probability tails that LES is not expected to simulate. The time averaged mean axial distribution of the vertical velocity also shows good agreement between the two setups.