DCPD and strain gauge based calibration procedure for evaluation of low temperature creep behavior

2021 ◽  
Vol 63 (7) ◽  
pp. 612-616
Author(s):  
Yinghao Cui ◽  
Zhang Jianlong ◽  
Xue He
Keyword(s):  
High Temperature ◽  
Creep Deformation ◽  
Nuclear Power ◽  
Water Environment ◽  
Gauge Length ◽  
Experimental Results ◽  
Theoretical Research ◽  
304 Stainless Steel ◽  
Primary Circuit ◽  
Service Water

Abstract Creep cracking is one of the key forms of structural material SCC damage with respect to nuclear power. Accurately obtaining the amount of creep deformation is also an important basis for estimating the service life of structural parts. However, because the primary circuit of nuclear power occurs in a high-temperature and high-pressure service water environment, it is not possible to use a conventional extensometer to obtain accurate creep of gauge length under these conditions. Considering that DCPD is an important method for monitoring crack propagation in a high-temperature water environment, by taking the austenite 304 stainless steel commonly used for nuclear power as a research object, a calibration method based on a combination of DC potential drop (DCPD) and strain testing to obtain the creep deformation of the specimen was established. By comparing theoretical research with experimental results, it can be concluded that the calculation results of the model are close to the experimental results and consistent with the theory, thus proving the feasibility of using DCPD technology to obtain the creep deformation amount.

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.

DEVELOPMENT OF THE HIGH TEMPERATURE FRETTING WEAR SIMULATOR FOR STEAM GENERATOR

2010 ◽  
Vol 24 (15n16) ◽  
pp. 2603-2608
Author(s):  
CHOON YEOL LEE ◽  
JOONG HO KIM ◽  
JOON WOO BAE ◽  
YOUNG SUCK CHAI
Keyword(s):  
High Temperature ◽  
Steam Generator ◽  
Nuclear Power ◽  
Contact Region ◽  
Water Environment ◽  
Control Unit ◽  
Fretting Wear ◽  
Test Rig ◽  
Realistic Condition ◽  
Wear Simulator

In nuclear power plant, fretting wear due to a combination of impact and sliding motions of the U-tubes against the supports and/or foreign objects caused by flow induced vibration, can make a serious problem in steam generator. A test rig, fretting wear simulator, is developed to elucidate fretting wear mechanism qualitatively and quantitatively. The realistic condition of steam generator of high temperature up to 320°C, high pressure up to 15 MPa, and water environment could be achieved by a test rig. The fretting wear simulator consists of main frame, water loop system, and control unit. Actual contact region under a realistic condition of steam generator was isolated using autoclave. Effects of various parameters such as the amounts of impact and sliding motions, applied loads and initial gaps and so forth are considered in this research. After the experiment, wear damage was measured by a three-dimensional profiler and the surface was also studied by SEM microscopically. Initial results were also presented.

Guided Waves As an Online Monitoring Technology for Long Term Operation in NPPs: New Experimental Results on a Steam Discharge Pipe

2017 ◽  
Author(s):  
Francesco Bertoncini ◽  
Mauro Cappelli ◽  
Francesco Cordella ◽  
Marco Raugi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Guided Waves ◽  
Complex Structure ◽  
Theoretical Background ◽  
Guided Wave ◽  
Experimental Results ◽  
Primary Circuit ◽  
Term Operation

Guided Wave (GW) testing is regularly used for finding defect locations through long range screening using low-frequency waves (from 5 to 250 kHz) [1]-[3]. Magnetostrictive sensors can overcome some issues, which usually limit the application to Nuclear Power Plants (NPPs) [4], like for example, high temperatures [5]-[6], high wall thickness of components in the primary circuit, and characteristic defect typologies. The authors have already shown the basic theoretical background, some simulations and some first experimental results concerning a real steel pipe, used for steam discharge, having a complex structure. Collecting more experimental data with a novel test campaign on the same pipe its complex structure results as a useful benchmark for the application of GWs as Non Destructive Techniques (NDT). Experimental measures using a symmetrical probe and a local probe in different configurations (pulse-echo and pitch-catch) indicate that GW testing with magnetostrictive sensors can be reliably applied to long-term monitoring of NPP components.

Research and Development of Compact Heat Exchangers of Gas Turbine Power Conversion Unit for Nuclear Power Plant With High Temperature Gas-Cooled Reactor

2010 ◽  
Author(s):  
V. F. Golovko ◽  
I. V. Dmitrieva ◽  
N. G. Kodochigov
Keyword(s):  
High Temperature ◽  
Heat Exchange ◽  
Gas Turbine ◽  
Nuclear Power ◽  
High Efficiency ◽  
Power Conversion ◽  
Experimental Investigations ◽  
Primary Circuit ◽  
Single Bundle ◽  
Conversion Unit

The NPP design that integrates a high temperature helium cooled nuclear reactor with a gas-turbine power conversion unit requires investigations and development of high-efficiency heat-exchange equipment operating in the closed primary circuit. The equipment must be very compact, which implies highly efficient heat transfer at minimum pressure loss. This paper presents an analysis of optimal heat-exchange surface selection, as well as design and layout features of recuperators, precoolers and intercoolers. Considered are tube (made of straight, helical, including those with the small bending radius, finned tubes etc.), plate-and-fin and matrix heat-exchange surfaces combined as separate modules or as a single bundle. Suggested are methods and criteria to select rational heat-exchange surfaces with account of critical factors and limitations. Given are results of the comparative analysis and computational and experimental investigations of surfaces; design and layout solutions for heat-exchange apparatuses arranged in the vertical high-pressure vessel with limited dimensions.

Experimental results of the Li-Pb-Pt system obtained by the high temperature drop calorimetry

2021 ◽  
pp. 115824
Author(s):  
S. Terlicka ◽  
A. Dębski ◽  
W. Gąsior ◽  
A. Fornalczyk ◽  
M. Saternus
Keyword(s):  
High Temperature ◽  
Temperature Drop ◽  
Experimental Results ◽  
Drop Calorimetry
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