Environmental Effects on Electrochemical Behaviors of Materials of a Dissimilar Metal Weld Exposed to High Temperature Primary Water of PWR Nuclear Power Plants

2015 ◽  
Vol 1088 ◽  
pp. 169-173
Author(s):  
Guang Fu Li ◽  
Ke Wei Fang ◽  
Jun Xu
Keyword(s):  
High Temperature ◽  
Dissolved Oxygen ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Electrochemical Behaviors ◽  
Dissimilar Metal ◽  
Primary Water ◽  
Dissimilar Metal Weld ◽  
Metal Weld

The effects of environmental factors on the electrochemical behaviors of the materials 52M and 316L taken from a dissimilar metal weld exposed to high temperature primary water of pressurized water reactor (PWR) nuclear power plants were studied experimentally, mainly on the effects of impurities chloride and sulfate in water, temperature and dissolved oxygen on the polarization curves, in order to provide fundamental data for relevant research and development. The results showed that doping chloride and sulfate into the water caused the rise of the tendency to pitting and general corrosion tendencies of both materials. With the rise of temperature from 160 °C to 290 °C, the tendencies to corrosion in anodic condition increased. The rise of the dissolved oxygen led to the rises of both the corrosion potentials and also the tendencies to corrosion.

Microstructures and Mechanical Properties of Dissimilar Metal Weld A508/52M/316L Used in Nuclear Power Plants

2011 ◽  
Vol 479 ◽  
pp. 40-47 ◽  
Author(s):  
Ke Wei Fang ◽  
Guan Jun Li ◽  
Guang Fu Li ◽  
Wu Yang ◽  
Mao Long Zhang ◽  
...  
Keyword(s):  
Mechanical Properties ◽  
Weld Metal ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Dissimilar Metal ◽  
Primary Water ◽  
Dissimilar Metal Weld ◽  
Microstructures And Mechanical Properties ◽  
Metal Weld

The microstructures and mechanical properties of a dissimilar metal weld A508/52M/316L used in the primary water system of pressurized water reactor (PWR) nuclear power plants were investigated. The weld exhibits complicated microstructures, with significant change around the interfaces A508/52M and 52M/316L. The variations of main elements in 52M weld metal are greater than those in the A508 and 316L, with significant changes in the zones closed to the interfaces. The bulk 52M weld metal has higher and more uneven hardness than both of the base metals A508 and 316L. The HAZ of A508 exhibits the highest hardness value in the weld. The area around the A508/52M interface is the most weak part for stress corrosion cracking (SCC) resistance of the weld in simulated PWR primary water at 290°C. SCC was only found in the specimens tested at +200mV(SHE) but not in those tested at both -780mV and Ecorr (about -500mV).

An Approach to Controlling Inaccessible Regions for Inspection Defined by Rules on Fitness-For-Service in Japan

2009 ◽  
Author(s):  
Takashi Ota ◽  
Koji Dozaki
Keyword(s):  
Ultrasonic Testing ◽  
Nuclear Power ◽  
Industrial Safety ◽  
Current Status ◽  
Inlet Pipe ◽  
Dissimilar Metal ◽  
Primary Water ◽  
Alternative Measures ◽  
Dissimilar Metal Weld ◽  
Metal Weld

On September 2007, Primary Water Stress Corrosion Cracking (PWSCC) flaws were found on the dissimilar metal weld of the steam generator (SG) inlet pipe nozzles of Tsuruga-2. Following the Tsuruga-2, similar cases were found in some other plants. These cracks were located in inaccessible regions by Ultrasonic Testing (UT) examination applied from the outer surface. Triggered by these cases, the Nuclear and Industrial Safety Agency (NISA), Japanese regulator of nuclear industries, directed nuclear power plant owners to investigate current status of these inaccessible regions for inspection defined by rules on fitness-for-service in Japan, and required to show developing plan on alternative measures of inspection for UT-exempted welds. On the other hand, the study to manage inaccessible regions in the rules on fitness-for-service has been started. The authors consider and propose a possible approach for modified rules of inspection in order to make control of these inaccessible regions.

Two- and Three-Dimensional Weld Residual Stress Analysis of a Weld Overlay Mockup of a VVER Feedwater Nozzle Dissimilar Metal Weld

2018 ◽  
Author(s):  
Francis H. Ku ◽  
Steven L. McCracken
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Pressurized Water ◽  
Weld Overlay ◽  
Dissimilar Metal ◽  
Before And After ◽  
Dissimilar Metal Weld ◽  
Three Dimensional Finite Element ◽  
Metal Weld

Weld overlay (WOL) is a popular repair technique to mitigate stress corrosion cracking (SCC) in dissimilar metal weld (DMW) in U.S. pressurized water reactor (PWR) design. The WOL technique is being considered as a SCC mitigation technique for DMW in Russian water-water energetic reactor (WWER or VVER) design. A WOL mockup on a VVER super emergency feedwater nozzle DMW has been fabricated, which represents the first WOL on VVER with the goal to mitigate SCC and the first WOL in Czech Republic civilian nuclear power plants. This paper presents the two- and three-dimensional finite element analyses performed to assess the weld residual stresses in the WOL mockup. The analysis evaluates the stress distribution and changes in the DMW before and after the WOL application, as well as compares the results to established industry guidelines and comparable WOLs on U.S. PWR.

scholarly journals Benefits, Drawbacks, and Future Trends of Brayton Helium Gas Turbine Cycles for Gas-Cooled Fast Reactor and Very-High Temperature Reactor Generation IV Nuclear Power Plants

2020 ◽  
Vol 7 (1) ◽  
Author(s):  
Arnold Gad-Briggs ◽  
Emmanuel Osigwe ◽  
Pericles Pilidis ◽  
Theoklis Nikolaidis ◽  
Suresh Sampath ◽  
...  
Keyword(s):  
High Temperature ◽  
Gas Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Optimum Operating ◽  
Future Trends ◽  
Design Point ◽  
Gen Iv ◽  
Very High

Abstract Numerous studies are on-going on to understand the performance of generation IV (Gen IV) nuclear power plants (NPPs). The objective is to determine optimum operating conditions for efficiency and economic reasons in line with the goals of Gen IV. For Gen IV concepts such as the gas-cooled fast reactors (GFRs) and very-high temperature reactors (VHTRs), the choice of cycle configuration is influenced by component choices, the component configuration and the choice of coolant. The purpose of this paper to present and review current cycles being considered—the simple cycle recuperated (SCR) and the intercooled cycle recuperated (ICR). For both cycles, helium is considered as the coolant in a closed Brayton gas turbine configuration. Comparisons are made for design point (DP) and off-design point (ODP) analyses to emphasize the pros and cons of each cycle. This paper also discusses potential future trends, include higher reactor core outlet temperatures (COT) in excess of 1000 °C and the simplified cycle configurations.

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