Cyclic Ramberg-Osgood Parameters of 316SS Weld Metal Under In-Air and PWR-Water Environments

2021 ◽  
Author(s):  
Jae Phil Park ◽  
Subhasish Mohanty ◽  
Chi Bum Bahn
Keyword(s):  
Room Temperature ◽  
Cyclic Plasticity ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Shakedown Analysis ◽  
Life Estimation ◽  
Fatigue Life Estimation ◽  
Tensile Curve ◽  
Asme Code ◽  
Primary Water

Abstract At present the available Ramberg-Osgood (R-O) parameters for different metals (e.g. in ASME code and other literature) are static (generally based on a tensile curve). These static R-O parameters cannot accurately model the cyclic plasticity behavior. This work presents the cyclic R-O material hardening parameters for 316 stainless steel similar metal welds. The parameters were estimated under various conditions (in-air at room temperature, 300°C in-air, and in-air at primary water conditions for a pressurized water reactor (PWR)). It is anticipated that the reported results would be useful for computational mechanics based shakedown analysis and fatigue life estimation of PWR components.

Examinations of Oxidation and Sulfidation of Grain Boundaries in Alloy 600 Exposed to Simulated Pressurized Water Reactor Primary Water

2013 ◽  
Vol 19 (3) ◽  
pp. 676-687 ◽  
Author(s):  
D.K. Schreiber ◽  
M.J. Olszta ◽  
D.W. Saxey ◽  
K. Kruska ◽  
K.L. Moore ◽  
...  
Keyword(s):  
Electron Microscopy ◽  
Grain Boundary ◽  
Grain Boundaries ◽  
Atom Probe Tomography ◽  
Atom Probe ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Intergranular Attack ◽  
Pressurized Water ◽  
Primary Water

AbstractHigh-resolution characterizations of intergranular attack in alloy 600 (Ni-17Cr-9Fe) exposed to 325°C simulated pressurized water reactor primary water have been conducted using a combination of scanning electron microscopy, NanoSIMS, analytical transmission electron microscopy, and atom probe tomography. The intergranular attack exhibited a two-stage microstructure that consisted of continuous corrosion/oxidation to a depth of ~200 nm from the surface followed by discrete Cr-rich sulfides to a further depth of ~500 nm. The continuous oxidation region contained primarily nanocrystalline MO-structure oxide particles and ended at Ni-rich, Cr-depleted grain boundaries with spaced CrS precipitates. Three-dimensional characterization of the sulfidized region using site-specific atom probe tomography revealed extraordinary grain boundary composition changes, including total depletion of Cr across a several nm wide dealloyed zone as a result of grain boundary migration.

Advanced Inlay System for Inlet/Outlet Nozzles of RV for Preventive Maintenance Against Alloy 600 PWSCC in Japanese PWR Plants

2013 ◽  
Author(s):  
Kazuhide Yamamoto ◽  
Masahiko Kizawa ◽  
Hiroki Kawazoe ◽  
Yuki Kobayashi ◽  
Ken Onishi ◽  
...  
Keyword(s):  
Preventive Maintenance ◽  
Heat Treatments ◽  
Residual Tensile Stress ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Nuclear Plants ◽  
Alloy 690 ◽  
Primary Water

Because many nuclear plants have been in operation for ages, the importance of preventive maintenance technologies is getting higher. One conspicuous problem found in pressurized water reactor (PWR) plants is the primary water stress corrosion cracking (PWSCC) observed in Alloy 600 (a kind of high nickel based alloy) parts. Alloy 600 was used for butt welds between low alloy steel and stainless steel of nozzles of Reactor Vessel (RV), Steam Generator (SG), and Pressurizer (Pz). As PWSCC occurred at these parts may cause Loss of Coolant Accident (LOCA), preventive maintenance is necessary. PWSCC is considered to be caused by a mixture of three elements: high residual tensile stress on surface, material (Alloy 600) and environment. PWSCC can be prevented by improving one of the elements. MHI has been developing stress improvement methods, for example, Water Jet Peening (WJP), Shot Peening by Ultrasonic vibration (USP), and Laser Stress Improvement Process (L-SIP). According to the situation, appropriate method is applied for each part. WJP has been applied for RV nozzles of a lot of plants in Japan. However PWSCC was observed in RV nozzles during the inspection before WJP in recent years, MHI developed the Advanced INLAY system to improve the material from Alloy 600 to Alloy 690. Alloy 600 on the inner surface of the nozzles is removed and welding with Alloy 690 is performed. In addition, heat treatments for the nozzles are difficult for its structural situation, so ambient temperature temper bead welding technique for RV nozzles was developed to make the heat treatments unnecessary. This paper describes the specifications of the advanced INLAY system and introduces the maintenance activities which MHI has applied for three plants in Japan by March 2012.

Kinetic Study of the Low Temperature Internal Oxidation of Nickel Based Model Alloys Exposed to PWR Primary Water

2008 ◽  
Vol 595-598 ◽  
pp. 449-462 ◽  
Author(s):  
Benoît Ter-Ovanessian ◽  
Julien Deleume ◽  
Jean Marc Cloué ◽  
Eric Andrieu
Keyword(s):  
Ternary Alloys ◽  
Pressurized Water Reactor ◽  
Oxide Interface ◽  
Intergranular Stress Corrosion ◽  
Pressurized Water ◽  
Free Oxygen ◽  
Primary Water ◽  
Nickel Base ◽  
The One ◽  
Intergranular Stress Corrosion Cracking

Two Ni-Fe-Cr ternary alloys have been oxidized in simulated pressurized water reactor primary water at 360°C for 1000 h. The chemical composition of those alloys were chosen in order to be representative of the one of chromium depleted areas under the oxide scale of industrial alloys (e.g. alloy 600) exposed in the same conditions. The resulting oxidized structures (corrosion scale and underlying metal) were characterized using complementary analytical methods (FEG-SEM, TEM, SIMS, optical microscopy). On the one hand, the characterized external oxide layer is very close to the one observed on industrial nickel-base alloys, hence validating the use of such model alloys. On the other hand, both free oxygen and oxides have been detected at grain boundaries several micrometers under the metal/oxide interface. Implications of such a finding on the involved transport mechanisms for oxygen and the intergranular stress corrosion cracking resistance of nickel-base alloys are then discussed.

Stress Corrosion Cracking Behavior of Alloy 182 Weld in Pressurized Water Reactor Primary Water Environment at 325ºC

CORROSION ◽  
2011 ◽  
Vol 67 (8) ◽  
pp. 085004-1-085004-9 ◽  
Author(s):  
L.I.L. Lima ◽  
M.M.A.M. Schvartzman ◽  
C.A. Figueiredo ◽  
A.Q. Bracarense
Keyword(s):  
Stress Corrosion ◽  
Oxide Layer ◽  
Stress Corrosion Cracking ◽  
Power Plants ◽  
Corrosion Cracking ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Cracking Behavior ◽  
Water Reactor ◽  
Primary Water

Abstract The weld used to connect two different metals is known as a dissimilar metal weld (DMW). In nuclear power plants, this weld is used to join stainless steel to low-alloy steel components in the nuclear pressurized water reactor (PWR). The most common weld metal is Alloy 182 (UNS W86182). Originally selected for its high corrosion resistance, it exhibited, after a long operation period, susceptibility to stress corrosion cracking (SCC) in PWR. The goal of this work was to study the electrochemical corrosion behavior and SCC susceptibility of Alloy 182 weld in PWR primary water containing 25 cm3 and 50 cm3 H2/kg H2O at standard temperature and pressure (STP). For this purpose, slow strain rate tensile (SSRT) tests and potentiodynamic polarization measurements were carried out. Scanning electron microscopy (SEM) with energy-dispersive spectrometry (EDS) was used to evaluate fracture morphology and determine the oxide layer chemical composition and morphology. The results indicated that at 325°C Alloy 182 weld is more susceptible to SCC at 25 cm3 (STP) H2/kg H2O and the increase of dissolved hydrogen decreased the crystal size of the oxide layer.

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