Dynamic Processes in Vacuum Contact Devices of Robots for Vertical Motion in the Water Environment

2019 ◽  
Vol 20 (7) ◽  
pp. 417-421
Author(s):  
V. G. Gradetsky ◽  
M. M. Knyazkov ◽  
E. A. Semenov ◽  
A. N. Sukhanov
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Water Environment ◽  
Experimental Studies ◽  
Vertical Motion ◽  
Contact Device ◽  
Dynamic Processes ◽  
Simplified Models ◽  
Contact Devices ◽  
Air Ejector

The results of experimental investigation intended to improve movement conditions for pneumatic robots on vertical surfaces under water are discussed. Features of the movement of vacuum contact devices for the simulation of mathematical model of the vacuum contact device with surfaces under water are presented. The experimental studies made it possible to obtain additional data on the dynamics of attachment, to obtain transient processes for air-water flow through ejector and to correct the results obtained earlier. For the purpose of analytical study of dynamic processes occurring in the system of vacuum contact devices, and taking into account the complexity of the description of nonlinearities, linearized simplified models of the system "air ejector — contact device — water environment" were developed. Vacuum contact devices are designed to provide guaranteed contact with vertical surfaces, plane slopes or horizontal surfaces on which the underwater robot performs its movement, carrying out the prescribed technological tasks, for example, in dry wells of nuclear power plants, on the surfaces of ship hulls, on the surfaces of underwater structures. The models took into account the forces of adhesion to the surfaces under water — the forces from the pressure drop, the friction force, the contact and vacuum interaction, the elasticity of suction caps. As a result of the solution of the model problem, the values of mechanical parameters, as well as the values of vacuum and flow in the cavity of variable volume as functions of changing the gap between the end of the corrugated membrane and the surfaces are obtained explicitly. As a result of the study of dynamic processes occurring in simplified models of vacuum contact devices "air ejector — contact surface — water environment", the transient characteristics of the change in the operating forces and pressures over time, as well as the dependence of the normal and tangential components of the forces on the depth of immersion in water were obtained. The variants of the designs of vacuum contact devices with surfaces in the water environment are investigated, and the modernization of the laboratory test bench for testing vacuum contact devices under water is carried out.

The Influence of Chemical Composition of Liquid Radioactive Waste on the Evaporation Process

2000 ◽  
Vol 663 ◽  
Author(s):  
P.P. Poluektov ◽  
L.P. Soukhanov ◽  
M.I. Zhicharev
Keyword(s):  
Nitric Acid ◽  
Chemical Composition ◽  
Radioactive Waste ◽  
Nuclear Power ◽  
Power Plants ◽  
Liquid Radioactive Waste ◽  
Salt Content ◽  
Evaporation Process ◽  
Simplified Models ◽  
Salt Systems

ABSTRACTA method is suggested to assess the tolerable salt content of the evaporator bottoms from the data on solubility in salt systems taken as simplified models of liquid radioactive waste (LRW) arising from nuclear power plants (NPP) with boiling reactors. It has been demonstrated that the degree of evaporation may be substantially increased by implementing the process in nitric acid. Equations have been derived that allow the calculation of the minimum needed acidity of the solution to allow maximum evaporation.

Over Limit Oxygen Effect on Crack Propagation of WWER 440 RPV Steel in Primary Circuit Water

2009 ◽  
Author(s):  
Miroslava Ernestova ◽  
Anna Hojna
Keyword(s):  
Crack Propagation ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Environment ◽  
Nuclear Research ◽  
Compact Tension ◽  
Oxygen Effect ◽  
Primary Circuit ◽  
Nuclear Research Institute ◽  
Primary Water

Experience with operating nuclear power plants worldwide reveals that many failures may be attributed to fatigue associated with mechanical loading due to vibration and with corrosion effect due to exposure to high-temperature environment. In order to clarify the simultaneous influence on reactor pressure vessel (RPV) material testing of ferritic steel 15Ch2MFA used for RPV of WWER 440 was performed at Nuclear Research Institute (NRI) autoclaves. Cyclic and constant loadings were applied to Compact Tension (CT) specimens in WWER primary water environment at 290°C and simultaneous effect of different oxygen levels (< 20 ppb, 200 ppb, 2000 ppb) on crack propagation has been evaluated. Obtained crack growth rates are compared with ASME XI Code and VERLIFE curves and crack behaviour is discussed.

Laser-Based Maintenance and Repair Technologies for Reactor Components

2004 ◽  
Author(s):  
Masaki Yoda ◽  
Naruhiko Mukai ◽  
Makoto Ochiai ◽  
Masataka Tamura ◽  
Satoshi Okada ◽  
...  
Keyword(s):  
Optical Fiber ◽  
Preventive Maintenance ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Rapid Heating ◽  
Water Environment ◽  
Inspection System ◽  
Material Surface ◽  
Maintenance And Repair

Stress corrosion cracking (SCC) is the major factor to reduce the reliability of aged reactor components. Toshiba has developed various laser-based maintenance and repair technologies and applied them to existing nuclear power plants. Laser-based technology is considered to be the best tool for remote processing in nuclear power plants, and particularly so for the maintenance and repair of reactor core components. Accessibility could be drastically improved by a simple handling system owing to the absence of reactive force against laser irradiation and the flexible optical fiber. For the preventive maintenance, laser peening (LP) technology was developed and applied to reactor components in operating BWR plants. LP is a novel process to improve residual stress from tensile to compressive on material surface layer by irradiating focused high-power laser pulses in water. We have developed a fiber-delivered LP system as a preventive maintenance measure against SCC. Laser ultrasonic testing (LUT) has a great potential to be applied to the remote inspection of reactor components. Laser-induced surface acoustic wave (SAW) inspection system was developed using a compact probe with a multi-mode optical fiber and an interferometer. The developed system successfully detected a micro slit of 0.5mm depth on weld metal and heat-affected zone (HAZ). An artificial SCC was also detected by the system. We are developing a new LP system combined with LUT to treat the inner surface of bottom-mounted instruments (BMI) of PWR plants. Underwater laser seal welding (LSW) technology was also developed to apply surface crack. LSW is expected to isolate the crack tip from corrosive water environment and to stop the propagation of the crack. Rapid heating and cooling of the process minimize the heat effect, which extends the applicability to neutron-irradiated material. This paper describes recent advances in the development and application of such laser-based technologies.

Evaluation of Low-Cycle Fatigue in Simulated PWR Environment

2006 ◽  
Vol 326-328 ◽  
pp. 1011-1014 ◽  
Author(s):  
Ill Seok Jeong ◽  
Sang Jai Kim ◽  
Taek Ho Song ◽  
Sung Yull Hong
Keyword(s):  
Fatigue Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Low Cycle Fatigue ◽  
Water Environment ◽  
Fatigue Tests ◽  
Cycle Fatigue ◽  
Fatigue Design ◽  
Number Of Cycles ◽  
Measured Length

For developing fatigue design curve of cast stainless steel that is used in piping material of nuclear power plants, a low-cycle fatigue test rig was built. It is capable of performing tests in pressurized high temperature water environment of PWR. Cylindrical solid fatigue specimens of CF8M were used for the strain-controlled environmental fatigue tests. Fatigue life was measured in terms of the number of cycles with the variation of strain amplitude at 0.04%/s strain rates. The disparity between target length and measured length of specimens was corrected by using finite element method. The corrected test results showed similar fatigue life trend with other previous results.

Structure and Mechanical Properties of Welded Joints for Nuclear Power Plants of Type MIR 1200

2017 ◽  
Vol 891 ◽  
pp. 201-205
Author(s):  
Ladislav Kander ◽  
Petr Čížek ◽  
Šárka Hermanová ◽  
Zdeněk Říha
Keyword(s):  
Mechanical Properties ◽  
Welded Joints ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Shear Fracture ◽  
Low Cycle Fatigue ◽  
Water Environment ◽  
Temperature Shift ◽  
Pressure Vessels

The paper deals with research, development and verification of production technology of selected welded joints for pressure vessels of primary circuits of nuclear power plants of type MIR 1200. Effect of various welding technology including simulation heat treatment on mechanical and fracture properties have been studied. Four type of homogenous 10GN2MFA – 10GN2MFA type of welded joints have been prepared for experimental programme. Conventional mechanical properties (tensile and impact test) as well as unconventional mechanical properties (fracture mechanics, low-cycle fatigue and stress corrosion cracking in water environment) have been studied. Effect of elevated working temperature on structure and material properties has been evaluated. Temperature dependencies of shear fracture have been plotted and effect of welding procedure on transition temperature shift has been evaluated. Experimental data have been compared with numerical simulation using FEM.

Experimental Studies of Radiation-Protective Properties of a Modified Titanium Hydride

2022 ◽  
Vol 1049 ◽  
pp. 174-179
Author(s):  
A.A. Karnauhov ◽  
R.N. Yastrebinskii
Keyword(s):  
Gamma Radiation ◽  
Dose Rate ◽  
Titanium Hydride ◽  
Nuclear Power ◽  
Power Plants ◽  
High Efficiency ◽  
Experimental Studies ◽  
Point Sources ◽  
Relaxation Length ◽  
Protective Properties

The results of experimental studies of the protective properties of titanium hydride with respect to neutron and gamma radiation in order to determine the optimal conditions for their use in the composition of the structural radiation protection of the nuclear reactor are presented. The weakening of the basic functionals in the thickness of protection, including the density of fast, intermediate and thermal neutrons, and the dose rate of gamma radiation is established. The functions of weakening the density of neutron flow and the dose rate of gamma radiation are measured in the conditions of "barrier" geometry. Determination of the protective properties of the structure was carried out when the modified titanium hydride fraction was placed in aluminum containers with a filling coefficient of a volume of container 0.63. The relaxation lengths for all neutron groups are close and on average are 9.8 cm. The functions of weakening the dose rate of gamma radiation of point sources Cs-137 and Co-60 are exponential. The weakening of radiation occurs with a constant relaxation length. For energy 0.661 MeV, the relaxation length is 7.1 cm, for energy 1.25 MeV, the relaxation length is equal to 10.1 cm. On the basis of the experimental studies, the high efficiency of the modified fraction of titanium hydride was confirmed during its use in protecting nuclear power plants.

Modelling of Pulse Height Spectra of Gamma Detectors for Simplified Models of a Radioactive Cloud

ANRI ◽  
2021 ◽  
Vol 0 (1) ◽  
pp. 45-52
Author(s):  
Sergey Gavrilov ◽  
Egor Il'ichev ◽  
Aleksey Kisilev ◽  
Artem Pimenov ◽  
Anton Shvedov
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pulse Height ◽  
Point Of View ◽  
Nuclear Industry ◽  
Simplified Models ◽  
Radioactive Cloud ◽  
Radionuclide Composition ◽  
Gamma Sources ◽  
Conducting Research

The paper considers the issues of determining the pulse height spectra of gamma detector from a radioactive cloud. This task is of interest from the point of view of possible improvement of existing systems for monitoring the radiation situation around nuclear power plants and nuclear industry enterprises due to the wider use of gamma-spectrometric equipment. Modeling of pulse height spectra will allow conducting research on the capabilities of monitoring system posts for detecting radionuclides in the radioactive cloud. A general approach to modeling pulse height spectra using division of the radioactive cloud into elementary gamma sources is developed. The pulse height spectra of scintillator NaI ∅63×63 mm are calculated for simplified models of the radioactive cloud in the form of a linear gamma source and a semi-infinite space. The obtained data can be used for rapid estimates of pulse height spectra, while the formulated approach to spectra modeling also allows for more time-consuming calculations for an arbitraryshaped radioactive cloud with an arbitrary radionuclide composition.

Research Dispersing Liquid and Gas in the Contact Device with an Increased Range of Stable Operation

2017 ◽  
Vol 21 (3) ◽  
pp. 12-15 ◽  
Author(s):  
A.V. Dmitriev ◽  
I.N. Madyshev ◽  
O.S. Dmitrieva ◽  
A.N. Nikolaev
Keyword(s):  
Mass Transfer ◽  
Heat And Mass Transfer ◽  
Distinctive Feature ◽  
Experimental Studies ◽  
Gas Bubbles ◽  
Contact Device ◽  
Stable Operation ◽  
Contact Devices ◽  
Transfer Apparatus ◽  
Axisymmetric Perturbations

To increase the efficiency of mass-transfer apparatus, a jet-bubbling contact device is suggested. The article considers perspectives for using of jet-bubble contact devices for heat and mass transfer apparatus. The distinctive feature of the developed device is an intensive countercurrent contact between the gas (vapor) and liquid in each element. The authors conducted experimental studies of dispersing liquid and gas in the proposed contact devices. Powerful turbulent axisymmetric perturbations occur in the bubbling layer, which affect the initial oscillations of the jet and determine the length of its decay. In the jets larger diameter distances between the involved gas bubbles less. Therefore, the resulting local jet are usually a smaller amount of liquid, which is decomposed accordingly into smaller drops.

Development of Underwater Laser Cladding and Underwater Laser Seal Welding Techniques for Reactor Components (III)

2008 ◽  
Author(s):  
Masataka Tamura ◽  
Yoshinobu Makino ◽  
Takehisa Hino ◽  
Shohei Kawano ◽  
Wataru Kouno ◽  
...  
Keyword(s):  
Laser Welding ◽  
Laser Cladding ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Environment ◽  
High Exposure ◽  
Maintenance And Repair ◽  
Underwater Laser ◽  
Repair Techniques

Recently, stress corrosion cracking (SCC) has been observed at aged components of nuclear power plants under water environment and high exposure of radiation. Toshiba has been developing both an underwater laser welding directly onto surface of the aged components as maintenance and repair techniques. This paper reports underwater laser cladding and seal welding using INCONEL 52/52M.

Study of a Gamma Radiation Impact on Concrete Properties Under Severe Accident Conditions

2020 ◽  
Vol 7 (2) ◽  
Author(s):  
Zbyněk Hlaváč ◽  
Jaroslava Zatloukalová ◽  
Michal Košťál ◽  
Evžen Losa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Radiation Effects ◽  
Experimental Studies ◽  
Life Time ◽  
Severe Accident ◽  
Research Centre ◽  
Concrete Properties ◽  
60Co Source ◽  
Operational Life

Abstract Concrete is an important structural material used in nuclear power plant (NPP) design. Due to relatively high amount of hydrogen as well as the presence of heavier elements, it also acts as a biological shielding. One of the important tasks for prolongation of operational life time is the determination of concrete components' condition after long-term irradiation. The paper aims to present the current activities in the CV Řež institute (Research Centre Řež—CVR) regarding the investigation of ionizing radiation effects on concrete properties. In its first part, the paper deals with experimental identification of the character of mixed neutron and gamma spectra in the concrete part of the VVER-1000 Mock-Up. Using the knowledge, the radiation field character can be scaled up to the commercial power plants with VVER-1000 light water reactor. It also provides justification for usage of the 60Co source for performed irradiation experiments with concrete. The second part of the article describes the experimental studies of the properties of gamma-irradiated concrete samples by strong 60Co source. This irradiation experiment can be understood as the first step in characterizing concrete degradation as gamma flux in biological shielding is significantly higher than that of neutron flux. In order to better understand the concrete properties and the behavior under irradiation, nondestructive as well as destructive testing methods were applied. We found that after 48 days of irradiation by the 60Co source the sample obtained dose from gamma corresponding to approximately 1% of the total during the NPP lifetime operation. Concrete microstructure degraded and the modulus of elasticity slightly decreased within 5%. Conversely, destruction tests prove significant flexural strength decrease by 27% in case of normal test and by 63% at the loss of coolant accident (LOCA) test.

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