scholarly journals Belarus Power Engineering System Modeling Taking into Account the Nuclear Power Plant Commissioning

2021 ◽  
Vol 64 (1) ◽  
pp. 5-14
Author(s):  
A. A. Mikhalevic ◽  
U. A. Rak
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Energy Systems ◽  
Operating Conditions ◽  
Power Engineering ◽  
Power Balance ◽  
Engineering System ◽  
Large Share ◽  
Load Duration

The article presents the analysis of the specific features of modeling the operation of energy systems with a large share of nuclear power plants (NPP). The study of operating conditions and characteristics of different power units showed that a power engineering system with a large share of NPP and CHPP requires more detailed modeling of operating modes of generating equipment. Besides, with an increase in the share of installations using renewable energy sources, these requirements are becoming tougher. A review of the literature revealed that most often the curve of the load duration and its distribution between blocks are used for modeling energy systems. However, since this method does not reflect a chronological sequence, it can only be used if there are no difficulties with ensuring power balance. Along with this, when the share of CHP and nuclear power plants is high, to maintain a balance of power one must know the parameters and a set of powered equipment not only currently but, also, in the previous period. But this is impossible if a curve of load duration is used. For modeling, it is necessary to use an hourly load curve and to calculate the state of the energy system for each subsequent hour in chronological order. In the course of a comparative analysis of available computer programs, it was not possible to identify a suitable model among the existing ones. The article presents a mathematical model developed by the authors, which makes us possible to simulate the operation of a power engineering system with a large share of NPP and CHPP while maintaining the power balance for each hour of the forecast period. Verification of the proposed model showed good accuracy of the methods used.

Diagnosis of Corrosion Processes in Nuclear Power Plants Secondary Piping Structures

2020 ◽  
Author(s):  
Koushik A. Manjunatha ◽  
Andrea Mack ◽  
Vivek Agarwal ◽  
David Koester ◽  
Douglas Adams
Keyword(s):  
Experimental Test ◽  
Classification Accuracy ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Conditions ◽  
Condition Based Maintenance ◽  
Support Vector ◽  
Maintenance Strategy ◽  
Test Bed

Abstract The current aging management plans of passive structures in nuclear power plants (NPPs) are based on preventative maintenance strategies. These strategies involve periodic, manual inspection of passive structures using nondestructive examination (NDE) techniques. This manual approach is prone to errors and contributes to high operation and maintenance costs, making it cost prohibitive. To address these concerns, a transition from the current preventive maintenance strategy to a condition-based maintenance strategy is needed. The research presented in this paper develops a condition-based maintenance capability to detect corrosion in secondary piping structures in NPPs. To achieve this, a data-driven methodology is developed and validated for detecting surrogate corrosion processes in piping structures. A scaled-down experimental test bed is developed to evaluate the corrosion process in secondary piping in NPPs. The experimental test bed is instrumented with tri-axial accelerometers. The data collected under different operating conditions is processed using the Hilbert-Huang Transformation. Distributional features of phase information among the accelerometers were used as features in support vector machine (SVM) and least absolute shrinkage and selection operator (LASSO) logistic regression methodologies to detect changes in the pipe condition from its baseline state. SVM classification accuracy averaged 99% for all models. LASSO classification accuracy averaged 99% for all models using the accelerometer data from the X-direction.

scholarly journals Current State of Safety of the Nuclear Power Plants, Prospect of the Development of Nuclear Power Engineering and Conceptual Issues of Development Strategy of Environmentally Friendly Nuclear Power

10.12737/4944 ◽  
2014 ◽  
Vol 3 (3) ◽  
pp. 60-73 ◽  
Author(s):  
Хвостова ◽  
Marina Khvostova ◽  
Острецов ◽  
Igor Ostretsov ◽  
Кузнецов ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Environmentally Friendly ◽  
Nuclear Power Engineering ◽  
Power Engineering ◽  
Stress Tests ◽  
Current State

The article considers current state of safety of nuclear power engineering. It presents a brief summary of stress-tests at nuclear power plants in the European Union and Russia. It reveals that the power on breeders shall not develop due to its low efficiency, high expenses and the risk of propagation of nuclear materials. Moreover, construction of plutonium processing production operations on nuclear power plant platforms with breeders, production of mixed uranium-plutonium nuclear fuel and synthesis ofamericium-241 in the spent nuclear fuel calls ecological safety into question. The article also addresses conceptual issues of creation of environmentally friendly nuclear power on the basis of nuclear relativistic technology. It is shown that such power shall not produce "bomb" materials and, therefore, will find extensive application around the world. Thereby the most challenging international problems of the present will be solved. The new nuclear power can become a basis for hydrogen production, which might solve practically all problems of mankind, including even food, by means of nuclear energy.

Analysis of the HTGR Proliferation Resistance

2008 ◽  
Author(s):  
Vitaly V. Petrunin ◽  
Nikolay G. Kodochigov ◽  
Yury P. Sukharev ◽  
Sergey L. Osipov ◽  
Elena V. Marova
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Material ◽  
Nuclear Power Engineering ◽  
Power Engineering ◽  
Nuclear Technology

Increased interest in development of nuclear power engineering, first of all in non-nuclear countries, puts an emphasis on the designing of small and medium nuclear power plants and determines the growth in nuclear technology export from countries with advanced nuclear industries. It accentuates the issue of reduction of the nuclear material proliferation risk, which was repeatedly raised on the national an international levels (materials of INPRO, GNEP, IAEA).

On-Site Evaluation of Bending Stresses at the Root of Small Bore Pipes by Accelerometric Measurements

2003 ◽  
Author(s):  
J. Guillou ◽  
L. Paulhiac
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Numerical Models ◽  
Operating Conditions ◽  
Strain Gauges ◽  
Bending Stress ◽  
Piping Systems ◽  
Bending Stresses ◽  
Site Evaluation

Several vibration-induced failures at the root of small bore piping systems occurred in French nuclear power plants in past years. The evaluation of the failure risk of the small bore pipes requires a fair estimation of the bending stress under operating conditions. As the use of strain gauges is too time-consuming in the environmental conditions of nuclear power plants, on-site acceleration measurements combined with numerical models are easier to handle. It still requires yet a large amount of updating work to estimate the stress in multi-span pipes with elbows and supports. The aim of the present study is to propose an alternate approach using two accelerometers to measure the local nozzle deflection, and an analytical expression of the bending stiffness of the nozzle on the main pipe. A first formulation is based on a static deformation assumption, thus allowing the use of a simple analog converter to get an estimation of the RMS value of the bending stress. To get more accurate results, a second method is based on an Euler Bernoulli deformation assumption: a spectral analyzer is then required to get an estimation of the spectrum of the bending stress. A better estimation of its RMS value is then obtained. An experimental validation of the methods based on strain gauges has been successfully performed.

Wear Scar Progression of Impact-Fretting at Elevated Temperature for Steam Generator Tubes in Nuclear Power Plants

2006 ◽  
Vol 326-328 ◽  
pp. 1251-1254 ◽  
Author(s):  
Chi Yong Park ◽  
Jeong Keun Lee
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Conditions ◽  
Wear Scar ◽  
Fretting Wear ◽  
Test Material ◽  
The Impact ◽  
Steam Generator Tubes

Fretting wear generated by flow induced vibration is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Understanding of tube wear characteristics is very important to keep the integrity of the steam generator tubes to secure the safety of the nuclear power plants. Experimental examination has been performed for the purpose of investigating the impact fretting. Test material is alloy 690 tube and 409 stainless steel tube supports. From the results of experiments, wear scar progression is investigated in the case of impact-fretting wear test of steam generator tubes under plant operating conditions such as pressure of 15MPa, high temperature of 290C and low dissolved oxygen. Hammer imprint that is actual damaged wear pattern, has been observed on the worn surface. From investigation of wear scar pattern, wear mechanism was initially the delamination wear due to cracking the hard oxide film and finally transferred to the stable impact-fretting pattern.

Crack Initiation Factor by Impact Fretting Wear at INCONEL Alloy for Steam Generator Tube in Nuclear Power Plants

2007 ◽  
Vol 26-28 ◽  
pp. 1269-1272
Author(s):  
Chi Yong Park ◽  
Jeong Kun Kim ◽  
Tae Ryong Kim ◽  
Sun Young Cho ◽  
Hyun Ik Jeon
Keyword(s):  
Steam Generator ◽  
Wear Mechanism ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Conditions ◽  
Fretting Wear ◽  
Steam Generator Tube ◽  
Inconel Alloy ◽  
The Impact

Inconel alloy such as alloy 600 and alloy 690 is widely used as the steam generator tube materials in the nuclear power plants. The impact fretting wear tests were performed to investigate wear mechanism between tube alloy and 409 stainless steel tube support plates in the simulated steam generator operating conditions, pressure of 15MPa, high temperature water of 290°C and low dissolved oxygen(<10 ppb). From investigation of wear test specimens by the SEM and EDS analysis, hammer imprint, which is known to be an actual damaged wear pattern, has been observed on the worn surface, and fretting wear mechanism was investigated. Wear progression of impact-fretting wear also has been examined. It was observed that titanium rich phase contributes to the formation of voids and cracks in sub-layer of fretting wear damage by impact fretting wear.

Investigation of the Fission Product Release From Molten Pools Under Oxidizing Conditions With the Code RELOS

2006 ◽  
Author(s):  
Ingo D. Kleinhietpaß ◽  
Hermann Unger ◽  
Hermann-Josef Wagner ◽  
Marco K. Koch
Keyword(s):  
Phase Boundary ◽  
Molten Pool ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fission Products ◽  
Energy Systems ◽  
The Core ◽  
Gas Atmosphere ◽  
Upper Pool

With the purpose of modeling and calculating the core behavior during severe accidents in nuclear power plants system codes are under development worldwide. Modeling of radionuclide release and transport in the case of beyond design basis accidents is an integrated feature of the deterministic safety analysis of nuclear power plants. Following a hypothetical, uncontrolled temperature escalation in the core of light water reactors, significant parts of the core structures may degrade and melt down under formation of molten pools, leading to an accumulation of large amounts of radioactive materials. The possible release of radionuclides from the molten pool provides a potential contribution to the aerosol source term in the late phase of core degradation accidents. The relevance of the amount of transferred oxygen from the gas atmosphere into the molten pool on the specification of a radionuclide and its release depends strongly on the initial oxygen inventory. Particularly for a low oxygen potential in the melt as it is the case for stratification when a metallic phase forms the upper layer and, respectively, when the oxidation has proceeded so far so that zirconium was completely oxidized, a significant influence of atmospheric oxygen on the specification and the release of some radionuclides has to be anticipated. The code RELOS (Release of Low Volatile Fission Products from Molten Surfaces) is under development at the Department of Energy Systems and Energy Economics (formerly Department of Nuclear and New Energy Systems) of the Ruhr-University Bochum. It is based on a mechanistic model to describe the diffusive and convective transport of fission products from the surface of a molten pool into a cooler gas atmosphere. This paper presents the code RELOS, i. e. the features and abilities of the latest code version V2.3 and the new model improvements of V2.4 and the calculated results evaluating the implemented models which deal with the oxygen transfer from the liquid side of the phase boundary to the bulk of the melt by diffusion or by taking into account natural convection. Both models help to estimate the amount of oxygen entering into the liquid upper pool volume and being available for the oxidation reaction. For both models the metallic, the oxidic and a mixture phase can be taken into account when defining the composition of the upper pool volume. The influence of crust formation, i. e. the decrease of the liquid pool surface area is taken care of because it yields the relevant amount of fission products released into the atmosphere. The difference of the partial density between the gas side of the phase boundary and the bulk of the gas phase is the driving force of mass transport.

scholarly journals IMPROVING GRINDING OF GEAR WHEELS APPLIED IN GEARBOXES OF POWER ENGINEERING

2021 ◽  
pp. 45-56
Author(s):  
Oleksiy Yakimov ◽  
Natalia Klimenko ◽  
Kateryna Kirkopulo ◽  
Andrey Pavlyshko ◽  
Sergyi Uminsky ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Operating Conditions ◽  
Power Engineering ◽  
Important Place ◽  
Continuous Increase ◽  
Method Of Calculation ◽  
Gear Grinding ◽  
Thermomechanical Processes

Development of modem power engineering follows the line of continuous increase in speed, coefficient of corrosive action and capacity of units. Gears and reducers are responsible parts of modem machinery and occupy an important place in the domestic power engineering construction. Durability and wear resistance of gears, apart from the design factors, also depends on the technological methods of treatment. The final stage of production of such wheels is the operation of gear grinding. In the process of gear grinding in a thin surface ball there are complex thermomechanical processes. As a result of short-time heating to high temperatures, structural transformations, burns, and in some cases even micro- and macro-thicknesses occur in such a surface bail. In addition, there are cases of making tooth wheels with adjacent defects grinding (for example, the appearance of the surface of the ball teeth of large tensioning forces), which reduces the life of the work, and in some cases causes a breakdown of the teeth in operating conditions. Development of effective measures to ensure the quality of the surface of the ball on the operation of grinding baggage in part depends on the possibility of predicting (or calculation) of temperatures and residual loads on the depth of the cemented teeth ball. The method of calculation of internal surplus Toads occurring during grinding of wheels with cemented steels is suggested. On the basis of the performed calculations and experiments the ways to improve the quality of production of working surfaces of gears, which are used in the wits of thermal and nuclear power plants are suggested and grounded.

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