scholarly journals Neutron flux simulation of a one-dimensional reactor zone taking into account xenon generation

Energetika ◽  
2020 ◽  
Vol 65 (4) ◽  
Author(s):  
Zsófia Tóth ◽  
Dániel Péter Kis
Keyword(s):  
Numerical Methods ◽  
Neutron Flux ◽  
Nuclear Power ◽  
Power Plants ◽  
Research Work ◽  
Reactor Core ◽  
Time And Space ◽  
One Dimensional ◽  
Safety And Reliability ◽  
The One

The energy in nuclear power plants is produced by thermal fission. It is extremely important to be able to monitor the processes in the reactor to ensure the safety and reliability of the power plant. One of the main traits of the reactor core is neutron flux. It changes in time and space therefore it is crucial to be able to simulate its changes with computer codes. In the research work a program code was established in the Matlab software with which the neutron flux of a one-dimensional zone can be simulated with homogenous and heterogenic zone parameters as well. The code is written using the one-group one-dimensional time- and space-dependent diffusion equation. The equation of an average delayed neutron group and xenon and iodine distributions was also included in the system to give a more precise look on the problem. The main innovation in the code is that numerical methods were used to solve the problem: the finite difference approach was applied for the place-dependent and for the time-dependent solution. The advantage of this code compared to other ones is that one-dimensional zones can be simulated in a really short time and it still gives a precise solution because of the complex numerical methods used.

scholarly journals INTELLIGENT TECHNIQUES FOR ANOMALY DETECTION IN NUCLEAR REACTORS

2021 ◽  
Vol 247 ◽  
pp. 21011
Author(s):  
George Ioannou ◽  
Thanos Tagaris ◽  
Georgios Alexandridis ◽  
Andreas Stafylopatis
Keyword(s):  
Signal Processing ◽  
Neutron Flux ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Machine Learning Techniques ◽  
Full Potential ◽  
Pressurized Water ◽  
Learning Techniques ◽  
Water Reactors

The safe operation of nuclear power plants is highly dependent on the ability of quickly and accurately identifying possible anomalies and perturbations in the reactor. Operational defects are primarily diagnosed by detectors that capture changes in the neutron flux, placed at various points inside and outside of the core. Neutron flux signals are subsequently analyzed with signal processing techniques in an effort to be better described (have their higher-order characteristics uncovered, locate transient events, etc). To this end, the application of intelligent techniques may be extremely beneficial, as it may assist and extend the current level of analysis. Besides, the combination of signal processing methodologies and machine learning techniques in the framework of nuclear power plant data is an emerging topic that has yet to show its full potential. In this context, the current contribution attempts at introducing intelligent approaches and more specifically, deep learning techniques, in neutron flux signal analysis for the identification of perturbations and other anomalies in the reactor core that may affect its operational capabilities. The obtained results of an initial stage of analysis on neutron flux signals captured at pressurized water reactors are encouraging, underlying the robustness and the potential of the proposed approach.

FMEDA-Based NPP I&C Systems Safety Assessment: Toward to Minimization of Experts’ Decisions Uncertainty

2016 ◽  
Author(s):  
Alexander Yasko ◽  
Eugene Babeshko ◽  
Vyacheslav Kharchenko
Keyword(s):  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Failure Modes ◽  
Electronic Components ◽  
Life Cycle Stages ◽  
Safety And Reliability ◽  
The One ◽  
And Control ◽  
Criticality Analysis

There are many techniques for the Nuclear Power Plants Instrumentation and Control (NPP I&C) systems safety assessment. Each of them is suitable for specific types of I&C systems and life cycle stages. Though general procedures of techniques application are specified by standards and described by guides, there is no universal solution that could be unambiguously applied to any NPP I&C system. The Failure Modes, Effects and Diagnostics/Criticality Analysis (FME(D/C)A) is the one that is most often used. Using this technique, the process of assessment is not trivial because of dimensionality problem that is especially critical for modern NPP I&C systems that contain many complex electronic components. Another challenge is the need of compliance to varying requirements of standards. Furthermore, modern I&C systems are based on different platforms (FPGA, microcontrollers). Most of safety and reliability assessment techniques, including mentioned FME(D/C)A, are based on expertise and thereby results are dependent on experts’ decisions very much. This could be a serious challenge, because it is very difficult to find universal experts that have sufficient experience to be equally qualified in different electronic components (FPGA, microcontrollers etc.) used in modern I&C systems. The goal of this paper is to analyze the ways of automation of FMEDA-based NPP I&C systems safety assessment and minimization of uncertainty degree of experts’ decisions. Possible experts’ errors and the uncertainty degree of their decisions are analyzed. We propose integration of all existing FMEA-based techniques into XME(D/C)A that includes Functional FMEA, Design FMEA, Software FMEA, Hardware FMEA etc. Such approach allows performing more comprehensive analysis. Developed tool AXMEA (Automated XMEA) represents an integrated solution that provides the automation of stages of FMEDA technique applied to NPP I&C. The case study is the application of proposed technique and tool during SIL3 certification of the modular RadICS™ platform.

Recent Patents on Valve Mechanism Device

2020 ◽  
Vol 13 (3) ◽  
pp. 230-241
Author(s):  
Ye Dai ◽  
Hui-Bing Zhang ◽  
Yun-Shan Qi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Flow Direction ◽  
Research Progress ◽  
Valve Mechanism ◽  
Advantages And Disadvantages ◽  
Valve Structure ◽  
Valve Leakage ◽  
Safety And Reliability

Background: Valves are an important part of nuclear power plants and are the control equipment used in nuclear power plants. It can change the cross-section of the passage and the flow direction of the medium and has the functions of diversion, cutoff, overflow, and the like. Due to the earthquake, the valve leaks, which will cause a major nuclear accident, endangering people's lives and safety. Objective: The purpose of this study is to synthesize the existing valve devices, summarize and analyze the advantages and disadvantages of various devices from many literatures and patents, and solve some problems of existing valves. Methods: This article summarizes various patents of nuclear-grade valve devices and recent research progress. From the valve structure device, transmission device, a detection device, and finally to the valve test, the advantages and disadvantages of the valve are comprehensively analyzed. Results: By summarizing the characteristics of a large number of valve devices, and analyzing some problems existing in the valves, the outlook for the research and design of nuclear power valves was made, and the planning of the national nuclear power strategic goals and energy security were planned. Conclusion: Valve damage can cause serious safety accidents. The most common is valve leakage. Therefore, the safety and reliability of valves must be taken seriously. By improving the transmission of the valve, the problems of complicated valve structure and high cost are solved.

Flow and Temperature Fields Measurement Inside Rod Bundle by the Combined Use of PIV and LIF Technique

2018 ◽  
Author(s):  
Xing Li ◽  
Sichao Tan ◽  
Zhengpeng Mi ◽  
Peiyao Qi ◽  
Yunlong Huang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Great Influence ◽  
Reactor Core ◽  
Transfer Characteristic ◽  
Index Of Refraction ◽  
Temperature Fields ◽  
Spacer Grid ◽  
Rod Bundle ◽  
Combined Use

Thermal hydraulic research of reactor core is important in nuclear energy applications, the flow and heat transfer characteristics of coolant in reactor fuel assembly has a great influence on the performance and safety of nuclear power plants. Particle image velocimetry (PIV) and Laser induced fluorescence (LIF) are the instantaneous, non-intrusive, whole-field fluid mechanics measuring method. In this study, the simultaneous measurement of flow field and temperature field for a rod bundle was conducted using PIV and LIF technique. A facility system, utilizing the matching index of refraction approach, has been designed and constructed for the measurement of velocity and temperature in the rod bundle. In order for further study on complex heat and mass transfer characteristic of rod bundle, the single-phase experiments on the heating conditions are performed. One of unique characteristics of the velocity and temperature distribution downstream the spacer grid was obtained. The experimental results show that the combined use of PIV and LIF technique is applied to the measurement of multi-physical field in a rod bundle is feasible, the measuring characteristics of non-intrusive ensured accuracy of whole field data. The whole field experimental data obtained in rod bundle benefits the design of spacer grid geometry.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

The analysis of emergency situations related to fires at nuclear power plants

2021 ◽  
Vol 30 (5) ◽  
pp. 66-75
Author(s):  
S. A. Titov ◽  
N. M. Barbin ◽  
A. M. Kobelev
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Emergency Situations ◽  
The Usa ◽  
Large Fires ◽  
Short Circuits ◽  
Nuclear Reactor Core

Introduction. The article provides a system and statistical analysis of emergency situations associated with fires at nuclear power plants (NPPs) in various countries of the world for the period from 1955 to 2019. The countries, where fires occurred at nuclear power plants, were identified (the USA, Great Britain, Switzerland, the USSR, Germany, Spain, Japan, Russia, India and France). Facilities, exposed to fires, are identified; causes of fires are indicated. The types of reactors where accidents and incidents, accompanied by large fires, have been determined.The analysis of major emergency situations at nuclear power plants accompanied by large fires. During the period from 1955 to 2019, 27 large fires were registered at nuclear power plants in 10 countries. The largest number of major fires was registered in 1984 (three fires), all of them occurred in the USSR. Most frequently, emergency situations occurred at transformers and cable channels — 40 %, nuclear reactor core — 15 %, reactor turbine — 11 %, reactor vessel — 7 %, steam pipeline systems, cooling towers — 7 %. The main causes of fires were technical malfunctions — 33 %, fires caused by the personnel — 30 %, fires due to short circuits — 18 %, due to natural disasters (natural conditions) — 15 % and unknown reasons — 4 %. A greater number of fires were registered at RBMK — 6, VVER — 5, BWR — 3, and PWR — 3 reactors.Conclusions. Having analyzed accidents, involving large fires at nuclear power plants during the period from 1955 to 2019, we come to the conclusion that the largest number of large fires was registered in the USSR. Nonetheless, to ensure safety at all stages of the life cycle of a nuclear power plant, it is necessary to apply such measures that would prevent the occurrence of severe fires and ensure the protection of personnel and the general public from the effects of a radiation accident.

scholarly journals Assessment of water hammer effects on boiling water nuclear reactor core dynamics

2007 ◽  
Vol 22 (1) ◽  
pp. 18-33 ◽  
Author(s):  
Anis Bousbia-Salah
Keyword(s):  
Pressure Wave ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Water Hammer ◽  
Pressure Wave Amplitude ◽  
Sensitivity Studies ◽  
Rapid Transients ◽  
Complex Phenomena

Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients. .

scholarly journals Analysis of numerical studies on the thermal-hydraulic and neutronic thermal-hydraulic stability of supercritical water reactors

2021 ◽  
Vol 7 (4) ◽  
pp. 311-318
Author(s):  
Artavazd M. Sujyan ◽  
Viktor I. Deev ◽  
Vladimir S. Kharitonov
Keyword(s):  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Negative Impact ◽  
Reactor Core ◽  
Resistance Coefficient ◽  
Coolant Flow ◽  
Water Reactor ◽  
Hydraulic Stability ◽  
Supercritical Water Reactor

The paper presents a review of modern studies on the potential types of coolant flow instabilities in the supercritical water reactor core. These instabilities have a negative impact on the operational safety of nuclear power plants. Despite the impressive number of computational works devoted to this topic, there still remain unresolved problems. The main disadvantages of the models are associated with the use of one simulated channel instead of a system of two or more parallel channels, the lack consideration for neutronic feedbacks, and the problem of choosing the design ratios for the heat transfer coefficient and hydraulic resistance coefficient under conditions of supercritical water flow. For this reason, it was decided to conduct an analysis that will make it possible to highlight the indicated problems and, on their basis, to formulate general requirements for a model of a nuclear reactor with a light-water supercritical pressure coolant. Consideration is also given to the features of the coolant flow stability in the supercritical water reactor core. In conclusion, the authors note the importance of further computational work using complex models of neutronic thermal-hydraulic stability built on the basis of modern achievements in the field of neutron physics and thermal physics.

scholarly journals Fractional Heat Conduction Models and Thermal Diffusivity Determination

2015 ◽  
Vol 2015 ◽  
pp. 1-9 ◽  
Author(s):  
Monika Žecová ◽  
Ján Terpák
Keyword(s):  
Heat Conduction ◽  
Thermal Diffusivity ◽  
Numerical Methods ◽  
Fractional Order ◽  
Experimental Verification ◽  
Heat Conduction Equation ◽  
Historical Overview ◽  
One Dimensional ◽  
The One ◽  
Fractional Heat Conduction

The contribution deals with the fractional heat conduction models and their use for determining thermal diffusivity. A brief historical overview of the authors who have dealt with the heat conduction equation is described in the introduction of the paper. The one-dimensional heat conduction models with using integer- and fractional-order derivatives are listed. Analytical and numerical methods of solution of the heat conduction models with using integer- and fractional-order derivatives are described. Individual methods have been implemented in MATLAB and the examples of simulations are listed. The proposal and experimental verification of the methods for determining thermal diffusivity using half-order derivative of temperature by time are listed at the conclusion of the paper.

scholarly journals Impact of Energy Slope Averaging Methods on Numerical Solution of 1D Steady Gradually Varied Flow

2015 ◽  
Vol 62 (3-4) ◽  
pp. 101-119 ◽  
Author(s):  
Wojciech Artichowicz ◽  
Dzmitry Prybytak
Keyword(s):  
Numerical Methods ◽  
Numerical Solution ◽  
Numerical Integration ◽  
Cross Sections ◽  
Flow Model ◽  
One Dimensional ◽  
Averaging Methods ◽  
Integration Formulas ◽  
Different Types ◽  
The One

AbstractIn this paper, energy slope averaging in the one-dimensional steady gradually varied flow model is considered. For this purpose, different methods of averaging the energy slope between cross-sections are used. The most popular are arithmetic, geometric, harmonic and hydraulic means. However, from the formal viewpoint, the application of different averaging formulas results in different numerical integration formulas. This study examines the basic properties of numerical methods resulting from different types of averaging.

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