Unanticipated Accidents Research Articles

Voice-Activated Emergency Response: Integrating Arduino Nicla Voice with MIT App Inventor for Elderly Care in Smart Homes

Nowadays, the average lifespan of people is increasing steadily. A multitude of elderly people have the tendency to reside in their own preferred dwellings. However, as a result of this, it might be too late to discover some elders' sudden abnormal situations. Therefore, it is a great concern for them to instantly and effectively deliver their distress signals to the outside world under consciousness, due to the fact that unanticipated accidents may happen all at once. This proposed system integrates the embedded system “Arduino Nicla Voice” in detection with MIT App Inventor in display on the user's interface and deploys Bluetooth Low Energy(BLE) as the communication channel. First, 8 different values of window overlap were tested and the best outcome was used in the upcoming validation method. The overall system performance in terms of accuracy, precision, recall, loss, and F1-score were examined with a 10-fold stratified cross validation. A Receiving Operating Characteristic(ROC) graph and Area Under the ROC Curve(AUC) based on predicted probability were exhibited to see the performance of each fold and the micro-average as well. Then, limitations were discussed so as to improve the robustness of this PERS system for future studies.

Comparison between electrical resistivity tomography and tunnel seismic prediction 303 methods for detecting the water zone ahead of the tunnel face: A case study

Abstract In deep tunnels with hydro-geological conditions, it is paramount to investigate the geological structure of the region before excavating a tunnel; otherwise, unanticipated accidents may cause serious damage and delay the project. The purpose of this study is to investigate the geological properties ahead of a tunnel face using electrical resistivity tomography (ERT) and tunnel seismic prediction (TSP) methods. During construction of the Nosoud Tunnel located in western Iran, ERT and TSP 303 methods were employed to predict geological conditions ahead of the tunnel face. In this article, the results of applying these methods are discussed. In this case, we have compared the results of the ERT method with those of the TSP 303 method. This work utilizes seismic methods and electrical tomography as two geophysical techniques are able to detect rock properties ahead of a tunnel face. This study shows that although the results of these two methods are in good agreement with each other, the results of TSP 303 are more accurate and higher quality. Also, we believe that using another geophysical method, in addition to TSP 303, could be helpful in making decisions in support of excavation, especially in complicated geological conditions.

Analytical model of high-voltage transmission line subjected to the downburst wind with rainfall

There have been many reports recently on unanticipated galloping and collapse accidents of tower-line systems due to downburst wind with rainfall. Although wet downburst is characterized by high-velocity wind with rainfall, very little research work is involved with galloping of high-voltage transmission line induced by the downburst wind with rainfall. Thus, this article proposes a preliminary theoretical study aiming to provide an analytical model of the high-voltage transmission line subjected to the downburst wind with rainfall to explain some phenomena observed from field measurements. Through wind or rain–wind tunnel experiments, we obtained aerodynamic characteristics of the high-voltage conductor with different yaw angles and rainfall rates. Considering the variations of several factors such as wind velocity, rainfall rate, yaw angle, and attack angle, the proposed analytical model was created by finite element method and central differences with the obtained aerodynamic coefficients of the high-voltage conductor. The theoretical results accord well with the experimental data. The analytical model enables better comprehension of the galloping of the high-voltage transmission line subjected to the downburst wind with rainfall.

Computational and Experimental Validation of a Passive Heat Removal System for NPP Containment with VVER-1200

The EUR-required passive system for protection against unanticipated accidents in nuclear reactors has been developed. This system and the computational results obtained using three-dimensional computer codes are described. The experimental data are compared with numerical experiments performed using three-dimensional hydrodynamic codes and the KUPOL-M lumped-parameter code. The experiments made it possible to refine the closure relations used in integral codes.

Numerical Modeling of Steam-Gas Flow in NPP Containment with VVER and Passive Heat Removal

The main elements of a mathematical model and test results for the specialized code PGS-TK to be used for calculating the flow of a multicomponent steam–gas mixture in the containment of a power-generating unit in NPP with passive heat-removal and volumetric and surface condensation of steam are presented. It is shown that owing to the modern mathematical models and numerical methods the code makes it possible to perform accurate calculations of the containment temperature and pressure for an unanticipated accident without using any adjustable parameters or specific experimental approximations.

KMS Stand: Large-Scale Model of NPP Containment with VVER

The main technical characteristics of the KMS stand as a large-scale model for NPP containment with VVER are presented. It is shown that this stand can be used to study three-dimensional heat and mass transfer processes in the containment atmosphere and to test systems and equipment used in NPP with VVER under close to anticipated and unanticipated accident conditions. In the future, with appropriate development work and additions to the stand it will be possible to study the rate of removal of aerosol particles having different chemical properties, perfect new designs of catalytic hydrogen recombiners and test equipment with the limiting parameters of the containment atmosphere.

Real-Time Calculation of the Accident at the Fukushima-1 NPP (Japan) Using the Sokrat Code

Some results of a numerical analysis of the serious unanticipated accident at the Fukishima-1 NPP are presented. The analysis was performed in March 2011 as part of the support given by Rosatom, including in making decisions about possible protection of the population in the Far East region of Russia. On the basis of calculations performed with the SOKRAT code, the state of the core during the first week and the possible reasons of the explosions in the first three power units were evaluated: the core of all three reactors melted and in course of its degradation from 950 to 1100 kg of hydrogen could have been formed followed by the formation of a burning mixture. The computational data on hydrogen emission due to oxidation of the melt by steam agree with experiments performed with the QUENCH and PARAMETR facilities. It can be asserted 1.5 years after the accident that post-accident studies confirm the main results of the real-time calculations.

Molecular Ion Mechanism of HI Formation in the First Loop

A model for the formation of hydrogen iodide gas directly in the channels of the first loop of a reactor during a serious accident initiated, by a source of radiation, in a mixture of vapors of fission products and water during radiative decay is described. The main parameters of the model are determined. It is shown that substantial conversion of cesium iodide vapor into hydrogen iodide is possible. Modeling the formation of different iodine constants in the first loop of a reactor in the event of a hypothetical unanticipated accident is a necessary condition for evaluating the consequences of radionuclides escaping into the room of the containment shell and radiation contamination of the environment [1]. For such accidents, aside from the formation of a large number of aerosols of fission products, a characteristic effect is the occurrence of different chemical reactions that can have a large effect on the formation of volatile forms of radionuclides and thereby determine their transport outside the plant’s con

Development of program modules with space-time kinetics for calculating unanticipated accidents in fast reactors

The first version of the complex computer code COREMELT-2D for calculating in a two-dimensional geometry the nonstationary neutron-physical and thermohydraulic processes taking account of coolant boiling, melting of fuel elements and mixing of melted components of a fast-reactor core has been developed. The code consists of a thermohydraulic module COREMELT and the neutron-physical module RADAR. COREMELT implements a multicomponent and multiphase thermohydraulics model in R–Z geometry in the porous body approximation. RADAR solves a multigroup neutron diffusion equation in R–Z and X–Y geometry. In the RADAR module, a modified TRIGEX program with a built-in CONSYST module, associated with the BNAB-93 group-constants library, is used to prepare the dynamical neutron constants. A test calculation of a serious accident due to loss of system and reliable power with all emergency protection organs of the reactor failing simultaneously is performed for the BN-1200 design using the complex program developed. The results are compared with previous calculations of the same accident performed using a COREMELT version with a point model of neutron kinetics.

철도차량 운전실의 작업환경 개선을 위한 시험 및 평가 정보시스템의 구축

Recent technological advances in the domestic railway industry have turned out to deliver a profound impact on the forefront of other industries as well. Furthermore, the introduction of the high-speed railway systems has made it possible to travel around our country in a day, thereby changing our life styles a lot. However, it has been well recognized that many factors can cause unanticipated accidents during the operation. One of the causes is known to be the human errors made by the train operators in locomotive cabs. The problem is getting more serious as the trains run at high speed. As such, the objective of this paper is on how to improve the work environment of the locomotive cabs. Our approach is based on the systems engineering methodology. Specifically, we first identify the context and scope of the problem to be solved. We then continue our effort in deriving the requirements set to accomplish the improvement. These results are utilized in constructing a computer-aided management system for test and evaluation intended for the improvement. The approach taken and the results obtained in this paper is expected to make a contribution on the route to keeping our nation's technologies on the competitive edges in the high-speed railway systems industry.

Investigation of aerosol kinetics in the protective envelope when analyzing accidental emissions at nuclear plants

The course of an unanticipated accident with serious damage to the core is characterized by the emission of a large quantity of radioactive materials into the first loop and then into the space inside the protective shell. The majority of the fission products enter the protective shell in the form of radioactive aerosols, whose leakage largely determines the consequences of a serious accident. This makes it necessary to model the dynamics of multicomponent aerosols taking account of the coagulation, change of their composition and mass as a result of the condensation-evaporation and dissolution of the gaseous components as well as the precipitation of the particles on different surfaces. A program for calculating the aerosol kinetic during the course of a serious accident is examined.

Modeling processes in the melt localization facility for aes-2006 with vver-1200 during an unanticipated accident

The results of a computational analysis of the behavior of melt in the facility used to localize melt for VVER-1200 (AES-2006 design), which were obtained using the GEFEST-ULR code, are presented. The main parameters of the melt are determined: the time variation of the component composition, the temperature and density of the oxide and metallic components and their relative arrangement. Using the KORSAR/GP thermohydraulic code, the minimum margin to crisis of heat transfer at the outer surface of the wall of the melt localization facility is analyzed. It is confirmed that heat transfer from the outer surface of the wall with normal operation of the cooling loop is reliable.

Investigation of the influence of the sodium void effect of reactivity on the technical-economic performance and safety of an advanced fast reactor

Measures to decrease the sodium void effect of reactivity and the influence of this effect on the technical-economic performance and the safety of BN-1200 are analyzed. Three variants of the core structure differing by the structural implementation and dimensions are examined. It is shown that a flattened core with a sodium cavity, replacing the top end screen, gives self-protection of the reactor with respect to unanticipated accidents. The elimination of the sodium cavity and an increase of the core height result in degradation of the self-protection properties but at the same time improve the technical-economic properties of the reactor. The possibilities for optimizing the construction of the reactor from the standpoint of reaching a compromise between safety properties and the technical-economic characteristics are discussed.

Unanticipated accidents in sodium-cooled fast reactors

The definitions and requirements of normative documents for unanticipated accidents at nuclear power plants with fast reactors are analyzed. Definitions are constructed between one another and with a collection of scenarios which can lead to unanticipated accidents, likewise determined by normative documents independently of the probability of these accidents actually happening. It is concluded that the normative approaches to fast-reactor safety must be refined with respect to strengthening the probabilistic criteria as a tool limiting the list of required unanticipated accidents for validating reactor safety. Special attention is devoted to the need to strengthen the motivation of designers to make the maximum possible use of passively triggered safety systems.

Experimental validation of the cooling loop for a passive system for removing heat from the AES-2006 protective envelope design for the Leningradskaya nuclear power plant site

Equipping new-generation nuclear power plants with passive means for controlling unanticipated accidents is one of the most promising directions for increasing safety, which is being implemented in the AES-2006 design for the site of the Leningradskaya nuclear power plant. An urgent problem is to obtain experimental validation of the passive system for removing heat from the protective envelope during unanticipated accidents with loss of coolant from the first loop in the case where the active systems fail. A particularity of the system is its state of constant readiness. The system functions with natural circulation of the coolant in both loops. Considering the importance of the passive heat removal system for ensuring the localizing properties of the protective envelope, OKBM Afrikantov has developed a large-scale stand and performed experimental investigations on validation of the effectiveness and serviceability of the cooling loop of a passive system for removing heat from the protective envelope.

Behavior of a model VVER fuel assembly under the conditions of an unanticipated accident with cooldown by pouring water from above

Investigations of the behavior of model VVER-1000 fuel assemblies have been performed on the PARAMETR bench under conditions of an unanticipated accident with reserve water poured from above. The results of materials engineering investigations of a model fuel assembly, using optical and electron microscopy, as well as x-ray microspectral and x-ray structural analysis are presented. The results of these investigations are necessary for developing and verifying methods of computing improved evaluations for the purpose of modeling accidents with substantial core damage.

Analysis of damage to RBMK core components as a result of an unanticipated accident

The results of a computational analysis of the damage to RBMK core components occurring during unanticipated accidents are presented. The deformation and scale of damage to fuel elements, non-bearing fuel-assembly rods, and channel pipes are examined for three accidents. The calculations show that all fuel elements become unsealed in the accidents examined. Substantial deformation of cladding, resulting in blocking of the flow section of a RBMK fuel channel, does not occur. The central carrying rod (tube) is likely to break with standard and central clamping of fuel element bundles in an assembly. Rupture of fuel channels is unavoidable in all unanticipated accidents studied.

Assessment of the systems efficiency of nuclear power plants with VVER taking account of regional factors

Methodological questions concerning the assessment of the efficiency and competitiveness of nuclear power plants with VVER taking account of regional factors are presented. The methodological and practical questions of how to increase the operational safety of nuclear power plants by controlling unanticipated accidents and increase the installed capacity utilization factor (ICUF) by forming and developing regional energy markets and by regional socio-economic development are presented.

Estimate of the emission rate of radioactive rare gases during an unanticipated accident at a VVÉR-1500 reactor

A method for estimating the emission rate of radioactive rare gases through the exhaust pipes of the passive filtration system in the intershell space of VVER-1500 and the ventilation pipes of the nuclear power plant is examined. It is proposed that flow and nonflow ionization chambers, which are located either in the exhaust pipes of the passive filtration system in the intershell space of the reactor or in the ventilation pipes of the nuclear power plant, and a spectrometric γ radiation sensor be used to obtain the estimate.

Fluid Structure Interaction Analysis on Wall Thinned Pipes

The wall thinning due to erosion, corrosion and flow accelerated corrosion is one of critical issues in nuclear industry. To secure against loss of integrity of pipes with a flaw, ASME Code Section III and Code Case N-597 etc have been used in design and operating stages, respectively. However, despite of their inherent conservatisms, it may reach unanticipated accidents due to degradation at local region. In this paper, a new evaluation scheme is suggested to estimate load-carrying capacities of wall thinned pipes. At first, computational fluid dynamics analyses employing steady-state and incompressible flow are carried out to determine pressure distributions in accordance with conveying fluid. Then, the discriminate pressures are applied as input condition of structural finite element analyses to calculate local stresses at the deepest point. A series of combined analyses were performed for different fluid flow velocities as well as d/t, Rm/t and l/t ratios. The efficiency of proposed scheme was proven from comparison with conventional analyses results and it is recommended to consider the fluid structure interaction effect for exact integrity evaluation.